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Sample records for waste immobilization estudo

  1. High-level-waste immobilization

    International Nuclear Information System (INIS)

    Crandall, J.L.

    1982-01-01

    Analysis of risks, environmental effects, process feasibility, and costs for disposal of immobilized high-level wastes in geologic repositories indicates that the disposal system safety has a low sensitivity to the choice of the waste disposal form

  2. Immobilized waste leaching

    International Nuclear Information System (INIS)

    Suarez, A.A.

    1989-01-01

    The main mechanism by which the immobilized radioactive materials can return to biosphere is the leaching due to the intrusion of water into the repositories. Some mathematical models and experiments utilized to evaluate the leaching rates in different immobilization matrices are described. (author) [pt

  3. Immobilization of organic liquid wastes

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.

    1985-01-01

    This report describes a portland cement immobilization process for the disposal treatment of radioactive organic liquid wastes which would be generated in a a FFTF fuels reprocessing line. An incineration system already on-hand was determined to be too costly to operate for the 100 to 400 gallons per year organic liquid. Organic test liquids were dispersed into an aqueous phosphate liquid using an emulsifier. A total of 109 gallons of potential and radioactive aqueous immiscible organic liquid wastes from Hanford 300 Area operations were solidified with portland cement and disposed of as solid waste during a 3-month test program with in-drum mixers. Waste packing efficiencies varied from 32 to 40% and included pump oils, mineral spirits, and TBP-NPH type solvents

  4. Immobilization of radioactive waste in glass matrices

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1978-01-01

    A promising process for long-term management of high-level radioactive waste is to immobilize the waste in a borosilicate glass matrix. Among the most important criteria characterizing the integrity of the large-scale glass-waste forms are that they possess good chemical stability (including low leachability), thermal stability, mechanical integrity, and high radiation stability. Fulfillment of these criteria ensures the maximum margin of safety of glass-waste products, following solidification, handling, transportation, and long-term storage

  5. A disposal centre for immobilized nuclear waste

    International Nuclear Information System (INIS)

    1980-02-01

    This report describes a conceptual design of a disposal centre for immobilized nuclear waste. The surface facilities consist of plants for the preparation of steel cylinders containing nuclear waste immobilized in glass, shaft headframe buildings and all necessary support facilities. The underground disposal vault is located on one level at a depth of 1000 m. The waste cylinders are emplaced into boreholes in the tunnel floors. All surface and subsurface facilities are described, operations and schedules are summarized, and cost estimates and manpower requirements are given. (auth)

  6. Process arrangement options for Defense waste immobilization

    International Nuclear Information System (INIS)

    1980-02-01

    Current plans are to immobilize the SRP high-level liquid wastes in a high integrity form. Borosilicate glass was selected in 1977 as the reference waste form and a mjaor effort is currently underway to develop the required technology. A large new facility, referred to as the Defense Waste Processing Facility (DWPF) is being designed to carry out this mission, with project authorization targeted for 1982 and plant startup in 1989. However, a number of other process arrangements or manufacturing strategies, including staging the major elements of the project or using existing SRP facilities for some functions, have been suggested in lieu of building the reference DWPF. This study assesses these various options and compares them on a technical and cost basis with the DWPF. Eleven different manufacturing options for SRP defense waste solidification were examined in detail. These cases are: (1) vitrification of acid waste at current generation rate; (2) vitrification of current rate acid waste and caustic sludge; (3 and 4) vitrification of the sludge portion of neutralized waste; (5) decontamination of salt cake and storage of concentrated cesium and strontium for later immobilization; (6) processing waste in a facility with lower capacity than the DWPF; (7) processing waste in a combination of existing and new facilities; (8) waste immobilization in H Canyon; (9) vitrification of both sludge and salt; (10) DWPF with onsite storage; (11) deferred authorization of DWPF

  7. Low Temperature Waste Immobilization Testing Vol. I

    Energy Technology Data Exchange (ETDEWEB)

    Russell, Renee L.; Schweiger, Michael J.; Westsik, Joseph H.; Hrma, Pavel R.; Smith, D. E.; Gallegos, Autumn B.; Telander, Monty R.; Pitman, Stan G.

    2006-09-14

    The Pacific Northwest National Laboratory (PNNL) is evaluating low-temperature technologies to immobilize mixed radioactive and hazardous waste. Three waste forms—alkali-aluminosilicate hydroceramic cement, “Ceramicrete” phosphate-bonded ceramic, and “DuraLith” alkali-aluminosilicate geopolymer—were selected through a competitive solicitation for fabrication and characterization of waste-form properties. The three contractors prepared their respective waste forms using simulants of a Hanford secondary waste and Idaho sodium bearing waste provided by PNNL and characterized their waste forms with respect to the Toxicity Characteristic Leaching Procedure (TCLP) and compressive strength. The contractors sent specimens to PNNL, and PNNL then conducted durability (American National Standards Institute/American Nuclear Society [ANSI/ANS] 16.1 Leachability Index [LI] and modified Product Consistency Test [PCT]) and compressive strength testing (both irradiated and as-received samples). This report presents the results of these characterization tests.

  8. Waste Treatment & Immobilization Plant Project

    Data.gov (United States)

    Federal Laboratory Consortium — In southeastern Washington State, Bechtel National, Inc. is designing, constructing and commissioning the world's largest radioactive waste treatment plant for the...

  9. Synroc tailored waste forms for actinide immobilization

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  10. Quality assessment of immobilized wastes

    International Nuclear Information System (INIS)

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

    1988-01-01

    A final repository concept for LLW and ILW is being studied in Brazil. It is thus now possible to assess in a systematic way the requirements on the waste packages in each step of the treatment, conditioning, storage, transport, disposal and the quality control procedure needed to show the requirements are fulfilled. The methodology to perform this assessment is discussed in this paper. The results of this methodology is proposed as basis for the licencing of the disposal of different waste packages in Brazil. (author) [pt

  11. Glass forms for immobilization of Hanford wastes

    International Nuclear Information System (INIS)

    Schulz, W.W.; Dressen, A.L.; Hobbick, C.W.; Babad, H.

    1975-03-01

    Approximately 140 million liters of solid salt cake (mainly NaNO 3 ), produced by evaporation of aged alkaline high-level liquid wastes, will be stored in underground tanks when the present Hanford Waste Management Program is completed in the early 1980's. At this time also, large volumes of various other solid radioactive wastes (sludges, excavated Pu-contaminated soil, and doubly encapsulated 137 CsCl and 90 SrF 2 ) will be stored on the Hanford Reservation. All these solid wastes can be converted to immobile silicate and aluminosilicate glasses of low water leachability by melting them at 1100 0 to 1400 0 C with appropriate amounts of basalt (or sand) and other glass-formers such as B 2 O 3 or CaO. Reviewed in this paper are formulations and other melt conditions used successfully in batch tests to make glasses from actual and synthetic wastes; leachability and other properties of these glasses show them to be satisfactory vehicles for immobilization of the Hanford wastes. (U.S.)

  12. The immobilization of organic liquid wastes

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.

    1986-01-01

    This report describes a portland cement immobilization process for the disposal treatment of radioactive organic liquid wastes which would be generated in a FFTF fuels reprocessing line. An incineration system already on-hand was determined to be too costly to operate for the 100 to 400 gallons per year organic liquid. Organic test liquids were dispersed into an aqueous phosphate liquid using an emulsifier. A total of 109 gallons of potential and radioactive aqueous immiscible organic liquid wastes from Hanford 300 Area operations were solidified with portland cement and disposed of as solid waste during a 3 month test program with in-drum mixers. Waste packing efficiencies varied from 32 to 40% and included pump oils, mineral spirits, and TBP-NPH type solvents

  13. Immobilization of wet solid wastes at nuclear power plants

    International Nuclear Information System (INIS)

    Neilson, R.M. Jr.

    1977-01-01

    Wet solid wastes are classified into four basic types: spent resins, filter sludges, evaporator concentrates, and miscellaneous liquids. Although the immobilization of wet solid wastes is primarily concerned with the incorporation of the waste with a solidification agent, there are a number of other discrete operations or subsystems involved in the treatment of these wastes that may affect the immobilized waste product. The immobilization process may be broken down into five basic operations: waste collection, waste pretreatment, solidification agent handling, mixing/packaging, and waste package handling. The properties of the waste forms that are ultimately shipped from the reactor site are primarily influenced by the methods utilized during the waste collection, waste pretreatment and mixing/packaging operations. The mixing/packaging (solidification) operation is perhaps the most important stage of the immobilization process. The basic solidification agent types are: absorbants, hydraulic cement, urea-formaldehyde, bitumen, and other polymer systems

  14. Study of immobilization of radioactive wastes in asphaltic matrices and elastomeric residues by using microwave technique; Estudo da imobilizacao de rejeitos radioativos em matrizes asfalticas e residuos elastomericos utilizando a tecnica de microondas

    Energy Technology Data Exchange (ETDEWEB)

    Caratin, Reinaldo Leonel

    2007-07-01

    In the present work, the technique of microwave heating was used to study the immobilization of low and intermediate activity level radioactive waste, such as spent ion exchange resin used to remove undesirable ions of primary circuits of refrigeration in water refrigerated nuclear reactors, and those used in chemical and radionuclide separation columns in the quality control of radioisotopes. Bitumen matrices reinforced with some kinds of rubber (Neoprene{sup R}, silicon and ethylene-vinyl-acetate), from production leftovers or scraps, were used for incorporation of radioactive waste. The samples irradiation was made in a home microwave oven that operates at a frequency of 2.450 MHZ with 1.000 W power. The samples were characterized by developing assays on penetration, leaching resistance, softening, flash and combustion points, thermogravimetry and optical microscopy. The obtained results were compatible with the pattern of matrices components, which shows that technique is a very useful alternative to conventional immobilization methods and to those kinds of radioactive waste. (author)

  15. Immobilization of fission products in phosphate ceramic waste forms

    International Nuclear Information System (INIS)

    Singh, D.

    1996-01-01

    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

  16. Immobilization of IFR salt wastes in mortar

    International Nuclear Information System (INIS)

    Fisher, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes from the fuel cycle of an integral fast reactor (IFR). The IFR is a sodium-cooled fast reactor with metal fuel. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500 degrees C. This cell has a cadmium anode and a liquid salt electrolyte. The salt will be a low-melting mixture of alkaline and alkaline earth chlorides. This paper discusses one method being considered for immobilizing this treated salt, to disperse it in a portland cement-base motar, which would then be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canisters where it will solidify into a strong, leach-resistant material

  17. Immobilization of IFR salt wastes in mortar

    International Nuclear Information System (INIS)

    Fischer, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes produced by the fuel cycles of Integral Fast Reactors (IFR). The IFR is a sodium-cooled fast reactor with metal alloy fuels. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500/degree/C. This cell has a liquid cadmium anode in which the fuels are dissolved and a liquid salt electrolyte. The salt will be a mixture of either lithium, potassium, and sodium chlorides or lithium, calcium, barium, and sodium chlorides. One method being considered for immobilizing the treated nontransuranic salt waste is to disperse the salt in a portland cement-base mortar that will be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canister-molds where it will solidify into a strong, leach-resistant material. The set times must be longer than a few hours to allow sufficient time for processing, and the mortar must reach a reasonable compressive strength (/approximately/7 MPa) within three days to permit handling. Because fission product heating will be high, about 0.6 W/kg for a mortar containing 10% waste salt, the effects of elevated temperatures during curing and storage on mortar properties must be considered

  18. Immobilization of industrial waste in cement–bentonite clay matrix

    Indian Academy of Sciences (India)

    Unknown

    Immobilization of industrial waste in cement–bentonite clay matrix. I B PLECAS* and S ... high structural integrity and minimizing the risk of escape by leaching. ..... Radioactive Waste Management and Nuclear Fuel Cycle 14. 195. Plecas I ...

  19. Waste immobilization process development at the Savannah River Plant

    International Nuclear Information System (INIS)

    Charlesworth, D.L.

    1986-01-01

    Processes to immobilize various wasteforms, including waste salt solution, transuranic waste, and low-level incinerator ash, are being developed. Wasteform characteristics, process and equipment details, and results from field/pilot tests and mathematical modeling studies are discussed

  20. High-level waste immobilization program: an overview

    International Nuclear Information System (INIS)

    Bonner, W.R.

    1979-09-01

    The High-Level Waste Immobilization Program is providing technology to allow safe, affordable immobilization and disposal of nuclear waste. Waste forms and processes are being developed on a schedule consistent with national needs for immobilization of high-level wastes stored at Savannah River, Hanford, Idaho National Engineering Laboratory, and West Valley, New York. This technology is directly applicable to high-level wastes from potential reprocessing of spent nuclear fuel. The program is removing one more obstacle previously seen as a potential restriction on the use and further development of nuclear power, and is thus meeting a critical technological need within the national objective of energy independence

  1. Nuclear waste immobilization in iron phosphate glasses

    International Nuclear Information System (INIS)

    Garcia, D.A.; Rodriguez, Diego A.; Menghini, Jorge E.; Bevilacqua, Arturo

    2007-01-01

    Iron-phosphate glasses have become important in the nuclear waste immobilization area because they have some advantages over silicate-based glasses, such as a lower processing temperature and a higher nuclear waste load without losing chemical and mechanical properties. Structure and chemical properties of iron-phosphate glasses are determined in terms of the main components, in this case, phosphate oxide along with the other oxides that are added to improve some of the characteristics of the glasses. For example, Iron oxide improves chemical durability, lead oxide lowers fusion temperature and sodium oxide reduces viscosity at high temperature. In this work a study based on the composition-property relations was made. We used different techniques to characterize a series of iron-lead-phosphate glasses with uranium and aluminium oxide as simulated nuclear waste. We used the Arquimedes method to determine the bulk density, differential temperature analysis (DTA) to determine both glass transition temperature and crystallization temperature, dilatometric analysis to calculate the linear thermal expansion coefficient, chemical durability (MCC-1 test) and X-ray diffraction (XRD). We also applied some theoretic models to calculate activation energies associated with the glass transition temperature and crystallization processes. (author)

  2. ALKALINE TREATMENT AND IMMOBILIZATION OF SECONDARY WASTE FROM WASTE INCINERATION

    Directory of Open Access Journals (Sweden)

    Dariusz Mierzwiński

    2017-04-01

    Full Text Available This paper regards the possibility of using geopolymer matrix to immobilize heavy metals present in ash and slag from combustion of waste. In the related research one used the fly ash from coal combustion in one Polish CHP plant and the waste from Polish incineration plants. It was studied if the above-named waste materials are useful in the process of alkali-activation. Therefore, three sets of geopolymer mixtures were prepared containing 60, 50 and 30% of ash and slag from the combustion of waste and fly ash combustion of sewage skudge. The remaining content was fly ash from coal combustion. The alkali-activation was conducted by means of 14M solution of NaOH and sodium water glass. The samples, whose dimensions were in accordance with the PN-EN 206-1 norm, were subjected to 75°C for 24h. According to the results, the geopolymer matrix is able to immobilize heavy metals and retain compressive strength resembling that of concrete.

  3. Treatment and immobilization of intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    Lerch, R.E.; Greenhalgh, W.O.; Partridge, J.A.; Richardson, G.L.

    1977-01-01

    This paper discusses a new program underway to develop and demonstrate treatment and immobilization technologies for intermediate level wastes (ILW) generated in the nuclear fuel cycle. Initial work has defined the sources, quantities and types of wastes which comprise ILW. Laboratory studies are underway to define treatment technologies for liquid ILW which contains volatile contaminants and to define immobilization parameters for the residues resulting from treatment of ILW. Immobilization agents initially being evaluated for the various residues include cement, urea-formaldehyde, and bitumen although other immobilization agents will be studied. The program also includes development of acceptable test procedures for the final immobilized products as well as development of proposed criteria for storage, transportation, and disposal of the immobilized ILW. 20 figures, 10 tables

  4. Alternatives generation and analysis report for immobilized low-level waste interim storage architecture

    Energy Technology Data Exchange (ETDEWEB)

    Burbank, D.A., Westinghouse Hanford

    1996-09-01

    The Immobilized Low-Level Waste Interim Storage subproject will provide storage capacity for immobilized low-level waste product sold to the U.S. Department of Energy by the privatization contractor. This report describes alternative Immobilized Low-Level Waste storage system architectures, evaluation criteria, and evaluation results to support the Immobilized Low-Level Waste storage system architecture selection decision process.

  5. DOE materials program supporting immobilization of radioactive wastes

    International Nuclear Information System (INIS)

    Oertel, G.K.; Scheib, W.S. Jr.

    1979-01-01

    A summary is presented of the DOE program for developing waste-form criteria, immobilization processes, and generation and evaluation of performance characterization data. Interrelationships are discussed among repository design, materials requirements, immobilization process definition, quality assurance, and risk analysis as part of the National Environmental Policy Act and regulatory processes

  6. Treatment and immobilization of intermediate-level radioactive wastes

    International Nuclear Information System (INIS)

    Lerch, R.E.; Greenhalgh, W.O.; Partridge, J.A.; Richardson, G.L.

    1979-01-01

    A new program underway at the Hanford Engineering Development Laboratory (HEDL) to develop and demonstrate treatment and immobilization technologies for intermediate-level wastes (ILW) generated in the nuclear fuel cycle is discussed. ILW are defined as those liquid and solid radioactive wastes, other than high-level wastes and fuel cladding hulls, that in packaged form have radiation dose readings greater than 200 millirem/hr at the packaged surface and 10 millirem/hr at three feet from the surface. The IAEA value of 10 4 Ci/m 3 for ILW defines the upper limit. For comparative purposes, reference is also made to certain aspects of low-level radioactive wastes (LLW). Initial work has defined the sources, quantities and types of wastes which comprise ILW. Because of the wide differences in composition (e.g., acids, salt solutions, resins and zeolites, HEPA filters, etc.) the wastes may require different treatments, particularly those wastes containing volatile contaminants. The various types of ILW have been grouped into categories amenable to similar treatment. Laboratory studies are underway to define treatment technologies for liquid ILW which contain volatile contaminants and to define immobilization parameters for the residues resulting from treatment of ILW. Immobilization agents initially being evaluated for the various residues include cement, urea-formaldehyde, and bitumen although other immobilization agents will be studied. The program also includes development of acceptable test procedures for the final immobilized products as well as development of proposed criteria for storage, transportation, and disposal of the immobilized ILW

  7. Study of the surface crystallization and resistance to dissolution of niobium phosphate glasses for nuclear waste immobilization; Estudo da cristalizacao superficial e da resistencia a dissolucao de vidros niobofosfatos visando a imobilizacao de rejeitos radioativos

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Heveline

    2008-07-01

    The surface crystallization and the dissolution rate of three phosphate glass compositions containing different amounts of niobium oxide were studied. The glasses were named Nb30, Nb37, and Nb44 according to the nominal content of niobium oxide in the glass composition. The three compositions were evaluated keeping the P{sub 2}O{sub 5}/K{sub 2}O ratio constant and varying the amount of Nb{sub 2}O{sub 5}. These glasses were produced by melting appropriate chemical compounds at 1500 deg C for 0.5 hour. The crystalline phases which were nucleated on the glass surface after heat treatment were determined by X-ray diffraction. The crystalline structures depend on the amount of niobium oxide in the glass composition. The crystal morphologies were observed by using an optical microscope, and their characteristics are specific for each kind of crystalline phase. The crystal growth rate and the surface nuclei density were determined for each glass composition, and they depend on each crystalline phase nucleated on the surface. From the differential thermal analysis curves it was determined that the Nb44 glass containing 46.5 mol por cent of niobium oxide is the most thermally stable against crystallization when compared to the Nb30 and Nb37 glasses. According to the activation energies determined for crystal growth on the surface of each glass type, the Nb44 glass can also be considered the most resistant one against crystallization. The dissolution rate for the Nb44 glass after 14 days immersed in an aqueous solution with pH equals to 7 at 90 deg C is the lowest (9.0 x 10{sup -7} g. cm{sup -2} . day{sup -1}) when compared to the other two glass compositions. The dissolution rates in acidic and neutral solutions of all studied glasses meet the international standards for materials which can be used in the immobilization of nuclear wastes. (author)

  8. Immobilization of hazardous and radioactive waste into glass structures

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1997-01-01

    As a result of more than three decades of international research, glass has emerged as the material of choice for immobilization of a wide range of potentially hazardous radioactive and non-radioactive materials. The ability of glass structures to incorporate and then immobilize many different elements into durable, high integrity, waste glass products is a direct function of the unique random network structure of the glassy state. Every major country involved with long-term management of high-level radioactive waste (HLW) has either selected or is considering glass as the matrix of choice for immobilizing and ultimately, disposing of the potentially hazardous, high-level radioactive material. There are many reasons why glass is preferred. Among the most important considerations are the ability of glass structures to accommodate and immobilize the many different types of radionuclides present in HLW, and to produce a product that not only has excellent technical properties, but also possesses good processing features. Good processability allows the glass to be fabricated with relative ease even under difficult remote-handling conditions necessary for vitrification of highly radioactive material. The single most important property of the waste glass produced is its ability to retain hazardous species within the glass structure and this is reflected by its excellent chemical durability and corrosion resistance to a wide range of environmental conditions. In addition to immobilization of HLW glass matrices are also being considered for isolation of many other types of hazardous materials, both radioactive as well as nonradioactive. This includes vitrification of various actinides resulting from clean-up operations and the legacy of the cold war, as well as possible immobilization of weapons grade plutonium resulting from disarmament activities. Other types of wastes being considered for immobilization into glasses include transuranic wastes, mixed wastes, contaminated

  9. Safe immobilization of high-level nuclear reactor wastes

    International Nuclear Information System (INIS)

    Ringwood, A.; Kesson, S.; Ware, N.; Hibberson, W.; Major, A.

    1979-01-01

    The advantages and disadvantages of methods of immobilizing high-level radioactive wastes are discussed. Problems include the devitrification of glasses and the occurrence of radiation damage. An alternative method of radioctive waste immobilization is described in which the waste is incorporated in the constituent minerals of a synthetic rock, Synroc. Synroc is immune from devitrification and is composed of phases which possess crystal structures identical to those of minerals which are known to have retained radioactive elements in geological environments at elevated pressures and tempertures for long periods. The composition and mineralogy of Synroc is given and the process of immobilizing wastes in Synroc is described. Accelerated leaching tests at elevated pressures and temperatures are also described

  10. Permitting plan for the immobilized low-activity waste project

    International Nuclear Information System (INIS)

    Deffenbaugh, M.L.

    1997-01-01

    This document addresses the environmental permitting requirements for the transportation and interim storage of the Immobilized Low-Activity Waste (ILAW) produced during Phase 1 of the Hanford Site privatization effort. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage and disposal of Tank Waste Remediation Systems (TWRS) immobilized low-activity tank waste (ILAW) and (2) interim storage of TWRS immobilized HLW (IHLW) and other canistered high-level waste forms. Low-activity waste (LAW), low-level waste (LLW), and high-level waste (HLW) are defined by the TWRS, Hanford Site, Richland, Washington, Final Environmental Impact Statement (EIS) DOE/EIS-0189, August 1996 (TWRS, Final EIS). By definition, HLW requires permanent isolation in a deep geologic repository. Also by definition, LAW is ''the waste that remains after separating from high-level waste as much of the radioactivity as is practicable that when solidified may be disposed of as LLW in a near-surface facility according to the NRC regulations.'' It is planned to store/dispose of (ILAW) inside four empty vaults of the five that were originally constructed for the Group Program. Additional disposal facilities will be constructed to accommodate immobilized LLW packages produced after the Grout Vaults are filled. The specifications for performance of the low-activity vitrified waste form have been established with strong consideration of risk to the public. The specifications for glass waste form performance are being closely coordinated with analysis of risk. RL has pursued discussions with the NRC for a determination of the classification of the Hanford Site's low-activity tank waste fraction. There is no known RL action to change law with respect to onsite disposal of waste

  11. Materials Science of High-Level Nuclear Waste Immobilization

    International Nuclear Information System (INIS)

    Weber, William J.; Navrotsky, Alexandra; Stefanovsky, S. V.; Vance, E. R.; Vernaz, Etienne Y.

    2009-01-01

    With the increasing demand for the development of more nuclear power comes the responsibility to address the technical challenges of immobilizing high-level nuclear wastes in stable solid forms for interim storage or disposition in geologic repositories. The immobilization of high-level nuclear wastes has been an active area of research and development for over 50 years. Borosilicate glasses and complex ceramic composites have been developed to meet many technical challenges and current needs, although regulatory issues, which vary widely from country to country, have yet to be resolved. Cooperative international programs to develop advanced proliferation-resistant nuclear technologies to close the nuclear fuel cycle and increase the efficiency of nuclear energy production might create new separation waste streams that could demand new concepts and materials for nuclear waste immobilization. This article reviews the current state-of-the-art understanding regarding the materials science of glasses and ceramics for the immobilization of high-level nuclear waste and excess nuclear materials and discusses approaches to address new waste streams

  12. Analysis of alternatives for immobilized low activity waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Burbank, D.A.

    1997-10-28

    This report presents a study of alternative disposal system architectures and implementation strategies to provide onsite near-surface disposal capacity to receive the immobilized low-activity waste produced by the private vendors. The analysis shows that a flexible unit strategy that provides a suite of design solutions tailored to the characteristics of the immobilized low-activity waste will provide a disposal system that best meets the program goals of reducing the environmental, health, and safety impacts; meeting the schedule milestones; and minimizing the life-cycle cost of the program.

  13. Analysis of alternatives for immobilized low-activity waste disposal

    International Nuclear Information System (INIS)

    Burbank, D.A.

    1997-01-01

    This report presents a study of alternative disposal system architectures and implementation strategies to provide onsite near-surface disposal capacity to receive the immobilized low-activity waste produced by the private vendors. The analysis shows that a flexible unit strategy that provides a suite of design solutions tailored to the characteristics of the immobilized low-activity waste will provide a disposal system that best meets the program goals of reducing the environmental, health, and safety impacts; meeting the schedule milestones; and minimizing the life-cycle cost of the program

  14. Description of processes for the immobilization of selected transuranic wastes

    International Nuclear Information System (INIS)

    Timmerman, C.L.

    1980-12-01

    Processed sludge and incinerator-ash wastes contaminated with transuranic (TRU) elements may require immobilization to prevent the release of these elements to the environment. As part of the TRU Waste Immobilization Program sponsored by the Department of Energy (DOE), the Pacific Northwest Laboratory is developing applicable waste-form and processing technology that may meet this need. This report defines and describes processes that are capable of immobilizing a selected TRU waste-stream consisting of a blend of three parts process sludge and one part incinerator ash. These selected waste streams are based on the compositions and generation rates of the waste processing and incineration facility at the Rocky Flats Plant. The specific waste forms that could be produced by the described processes include: in-can melted borosilicate-glass monolith; joule-heated melter borosilicate-glass monolith or marble; joule-heated melter aluminosilicate-glass monolith or marble; joule-heated melter basaltic-glass monolith or marble; joule-heated melter glass-ceramic monolith; cast-cement monolith; pressed-cement pellet; and cold-pressed sintered-ceramic pellet

  15. Hanford Immobilized Low-Activity Waste Product Acceptance Test Plan

    International Nuclear Information System (INIS)

    Peeler, D.

    1999-01-01

    'The Hanford Site has been used to produce nuclear materials for the U.S. Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during Pu production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The DOE is proceeding with an approach to privatize the treatment and immobilization of Handord''s LAW and HLW.'

  16. Hanford Immobilized Low-Activity Waste Product Acceptance Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, D.

    1999-06-22

    'The Hanford Site has been used to produce nuclear materials for the U.S. Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during Pu production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The DOE is proceeding with an approach to privatize the treatment and immobilization of Handord''s LAW and HLW.'

  17. Evaluation of bitumens for nuclear facilities radioactive waste immobilization

    International Nuclear Information System (INIS)

    Guzella, Marcia F.R.; Silva, Tania V. da; Loiola, Roberto; Monte, Lauro J.B.

    2000-01-01

    The activities developed at the Nuclear Technology Development Centre, Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN, include the research and development work of the radioactive wastes immobilization in different kind of bitumen. The present work describes the bituminization of simulated low level wastes of evaporator concentrates.Two types of bitumen are used for incorporation of the simulated wastes generated by nuclear power plants. Studies on rheological properties, leaching data, differential thermoanalysis and water content of the waste-products have been carried out. (author)

  18. The immobilization of High Level Waste Into Glass

    International Nuclear Information System (INIS)

    Aisyah; Martono, H.

    1998-01-01

    High level liquid waste is generated from the first step extraction in the nuclear fuel reprocessing. The waste is immobilized with boro-silicate glass. A certain composition of glass is needed for a certain type of waste, so that the properties of waste glass would meet the requirement either for further process or for disposal. The effect of waste loading on either density, thermal expansion, softening point and leaching rate has been studied. The composition of the high level liquid waste has been determined by ORIGEN 2 and the result has been used to prepare simulated high level waste. The waste loading in the waste glass has been set to be 19.48; 22.32; 25.27; and 26.59 weight percent. The result shows that increasing the waste loading has resulted in the higher density with no thermal expansion and softening point significant change. The increase in the waste loading increase that leaching rate. The properties of the waste glass in this research have not shown any deviation from the standard waste glass properties

  19. Utilization of immobilized urease for waste water treatment

    Science.gov (United States)

    Husted, R. R.

    1974-01-01

    The feasibility of using immobilized urease for urea removal from waste water for space system applications is considered, specifically the elimination of the urea toxicity problem in a 30-day Orbiting Frog Otolith (OFO) flight experiment. Because urease catalyzes the hydrolysis of urea to ammonia and carbon dioxide, control of their concentrations within nontoxic limits was also determined. The results of this study led to the use of free urease in lieu of the immobilized urease for controlling urea concentrations. An ion exchange resin was used which reduced the NH3 level by 94% while reducing the sodium ion concentration only 10%.

  20. Phase 1 immobilized low-activity waste operational source term

    International Nuclear Information System (INIS)

    Burbank, D.A.

    1998-01-01

    This report presents an engineering analysis of the Phase 1 privatization feeds to establish an operational source term for storage and disposal of immobilized low-activity waste packages at the Hanford Site. The source term information is needed to establish a preliminary estimate of the numbers of remote-handled and contact-handled waste packages. A discussion of the uncertainties and their impact on the source term and waste package distribution is also presented. It should be noted that this study is concerned with operational impacts only. Source terms used for accident scenarios would differ due to alpha and beta radiation which were not significant in this study

  1. Immobilized low-level waste disposal options configuration study

    International Nuclear Information System (INIS)

    Mitchell, D.E.

    1995-02-01

    This report compiles information that supports the eventual conceptual and definitive design of a disposal facility for immobilized low-level waste. The report includes the results of a joint Westinghouse/Fluor Daniel Inc. evaluation of trade-offs for glass manufacturing and product (waste form) disposal. Though recommendations for the preferred manufacturing and disposal option for low-level waste are outside the scope of this document, relative ranking as applied to facility complexity, safety, remote operation concepts and ease of retrieval are addressed

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

    International Nuclear Information System (INIS)

    Crisler, L.R.

    1980-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  5. Characterization plan for the immobilized low-activity waste borehole

    International Nuclear Information System (INIS)

    Reidel, S.P.; Reynolds, K.D.

    1998-03-01

    The US Department of Energy's (DOE's) Hanford Site has the most diverse and largest amounts of radioactive tank waste in the US. High-level radioactive waste has been stored at Hanford in large underground tanks since 1944. Approximately 209,000 m 3 (54 Mgal) of waste are currently stored in 177 tanks. Vitrification and onsite disposal of low activity tank waste (LAW) are embodied in the strategy described in the Tri-Party Agreement. The tank waste is to be retrieved, separated into low- and high-level fractions, and then immobilized by private vendors. The DOE will receive the vitrified waste from private vendors and dispose of the low-activity fraction in the Hanford Site 200 East Area. The Immobilized Low-Activity Waste Disposal Complex (ILAWDC) is part of the disposal complex. This report is a plan to drill the first characterization borehole and collect data at the ILAWDC. This plan updates and revises the deep borehole portion of the characterization plan for the ILAWDC by Reidel and others (1995). It describes data collection activities for determining the physical and chemical properties of the vadose zone and the saturated zone at and in the immediate vicinity of the proposed ILAWDC. These properties then will be used to develop a conceptual geohydrologic model of the ILAWDC site in support of the Hanford ILAW Performance Assessment

  6. Thermal process for immobilization of radioactive wastes

    International Nuclear Information System (INIS)

    Brownell, L.E.; Isaacson, R.E.; Kupfer, M.J.; Schulz, W.W.

    1971-01-01

    The Thermalt process involves an exothermic, thermite-like reaction of aluminum metal with basalt, quartz sand, and radioactive waste. The resulting melt when solidified is a silicious stone-like material that is similar in chemical composition to basalt. The process utilizes low cost ingredients: basalt rock, which occurs naturally in the Hanford region, inexpensive aluminum metal such as aluminum scrap which need not be pure, and the waste which is predominately sodium nitrate salt. The waste itself along with the basalt provides the oxygen necessary for the reaction. The exothermic reaction provides the necessary heat to melt the ingredients thus eliminating the need for external heat sources such as furnaces which are necessary with most other melt methods. The final product is highly stable and essentially nonleachable; leach rates appear as low or lower than other melt products described in the literature. Initial studies indicate the process is effective for both low-level and high-level wastes. (U.S.)

  7. Radioactive wastes immobilization in glasses and ceramics

    International Nuclear Information System (INIS)

    Zanotto, E.D.

    1983-01-01

    A review on the several options available for encapsulation and disposal of high level radioactive liquid wastes is presented, along with the relative merits and disadvantages of each material to be encapsulated. Some of the main fields requiring further advancements in both scientific and technological research are discussed and a few suggestions for the solution of the brazilian problem are given. (Author) [pt

  8. Bulk Vitrification Technology For The Treatment And Immobilization Of Low-Activity Waste

    International Nuclear Information System (INIS)

    Ard, K.E.

    2011-01-01

    This report is one of four reports written to provide background information regarding immobilization technologies under consideration for supplemental immobilization of Hanford's low-activity waste. This paper is intended to provide the reader with general understanding of Bulk Vitrification and how it might be applied to immobilization of Hanford's low-activity waste.

  9. Fluidized Bed Steam Reforming For Treatment And Immobilization Of Low-Activity Waste

    International Nuclear Information System (INIS)

    Hewitt, W.M.

    2011-01-01

    This report is one of four reports written to provide background information regarding immobilization technologies remaining under consideration for supplemental immobilization of Hanford's low-activity waste. This paper provides the reader a general understanding of fluidized bed steam reforming and its possible application to treat and immobilize Hanford low-activity waste.

  10. BULK VITRIFICATION TECHNOLOGY FOR THE TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE

    Energy Technology Data Exchange (ETDEWEB)

    ARD KE

    2011-04-11

    This report is one of four reports written to provide background information regarding immobilization technologies under consideration for supplemental immobilization of Hanford's low-activity waste. This paper is intended to provide the reader with general understanding of Bulk Vitrification and how it might be applied to immobilization of Hanford's low-activity waste.

  11. FLUIDIZED BED STEAM REFORMING FOR TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE

    Energy Technology Data Exchange (ETDEWEB)

    HEWITT WM

    2011-04-08

    This report is one of four reports written to provide background information regarding immobilization technologies remaining under consideration for supplemental immobilization of Hanford's low-activity waste. This paper provides the reader a general understanding of fluidized bed steam reforming and its possible application to treat and immobilize Hanford low-activity waste.

  12. Immobilization and Waste Form Product Acceptance for Low Level and TRU Waste Forms

    International Nuclear Information System (INIS)

    Holtzscheiter, E.W.; Harbour, J.R.

    1998-05-01

    The Tanks Focus Area is supporting technology development in immobilization of both High Level (HLW) and Low Level (LLW) radioactive wastes. The HLW process development at Hanford and Idaho is patterned closely after that of the Savannah River (Defense Waste Processing Facility) and West Valley Sites (West Valley Demonstration Project). However, the development and options open to addressing Low Level Waste are diverse and often site specific. To start, it is important to understand the breadth of Low Level Wastes categories

  13. Model for acquiring innovative waste immobilization technologies

    International Nuclear Information System (INIS)

    Dole, L.R.; Singh, S.P.N.

    1988-01-01

    The US Department of Energy's (DOE's) Oak Ridge Operations (ORO) has established the Waste Management Technology Center (WMTC) at Oak Ridge National Laboratory to assist in meeting the environmental requirements for federal facilities as stated in the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). The WMTC will bring innovative mixed chemical and radioactive waste treatment and site closure technologies to bear on the many mixed chemical and radioactive waste problems at the DOE-ORO facilities located in Tennessee, Kentucky, and Ohio. The WMTC seeks innovative technologies through a phased procurement cycle that encourages the teaming of emerging technologies with experienced contractors in order to comply with on-site requirements of DOE orders concerning protection of the environment. This three-phase procurement cycle includes: (1) a feasibility study and implementation plan, (2) an on-site pilot demonstration, and (3) full-scale implementation. This paper describes the statements of work for some related demonstrations and remedial actions

  14. ORNL grouting technologies for immobilizing hazardous wastes

    International Nuclear Information System (INIS)

    Dole, L.R.; Trauger, D.B.

    1983-01-01

    The Cement and Concrete Applications Group at the Oak Ridge National Laboratory (ORNL) has developed versatile and inexpensive processes to solidify large quantities of hazardous liquids, sludges, and solids. By using standard off the shelf processing equipment, these batch or continuous processes are compatible with a wide range of disposal methods, such as above-ground storage, shallow-land burial, deep geological disposal, sea-bed dumping, and bulk in-situ solidification. Because of their economic advantages, these latter bulk in-situ disposal scenarios have received the most development. ORNL's experience has shown that tailored cement-based formulas can be developed which tolerate wide fluctuations in waste feed compositions and still maintain mixing properties that are compatible with standard equipment. In addition to cements, these grouts contain pozzolans, clays and other additives to control the flow properties, set-times, phase separations and impacts of waste stream fluctuation. The cements, fly ashes and other grout components are readily available in bulk quantities and the solids-blends typically cost less than $0.05 to 0.15 per waste gallon. Depending on the disposal scenario, total disposal costs (material, capital, and operating) can be as low as $0.10 to 0.50 per gallon

  15. Immobilization of copper flotation waste using red mud and clinoptilolite.

    Science.gov (United States)

    Coruh, Semra

    2008-10-01

    The flash smelting process has been used in the copper industry for a number of years and has replaced most of the reverberatory applications, known as conventional copper smelting processes. Copper smelters produce large amounts of copper slag or copper flotation waste and the dumping of these quantities of copper slag causes economic, environmental and space problems. The aim of this study was to perform a laboratory investigation to assess the feasibility of immobilizing the heavy metals contained in copper flotation waste. For this purpose, samples of copper flotation waste were immobilized with relatively small proportions of red mud and large proportions of clinoptilolite. The results of laboratory leaching demonstrate that addition of red mud and clinoptilolite to the copper flotation waste drastically reduced the heavy metal content in the effluent and the red mud performed better than clinoptilolite. This study also compared the leaching behaviour of metals in copper flotation waste by short-time extraction tests such as the toxicity characteristic leaching procedure (TCLP), deionized water (DI) and field leach test (FLT). The results of leach tests showed that the results of the FLT and DI methods were close and generally lower than those of the TCLP methods.

  16. Immobilization of high-level wastes into sintered glass: 1

    International Nuclear Information System (INIS)

    Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

    1987-01-01

    In order to immobilize the high-level radioactive wastes from fuel elements reprocessing, borosilicate glass was adopted. Sintering experiments are described with the variety VG 98/12 (SiO 2 , TiO 2 , Al 2 O 3 , B 2 O 3 , MgO, CaO and Na 2 O) (which does not present devitrification problems) mixed with simulated calcinated wastes. The hot pressing line (sintering under pressure) was explored in two variants 1: In can; 2: In graphite matrix with sintered pellet extraction. With scanning electron microscopy it is observed that the simulated wastes do not disolve in the vitreous matrix, but they remain dispersed in the same. The results obtained point out that the leaching velocities are independent from the density and from the matrix type employed, as well as from the fact that the wastes do no dissolve in the matrix. (M.E.L.) [es

  17. Hanford immobilized low-activity tank waste performance assessment

    International Nuclear Information System (INIS)

    Mann, F.M.

    1998-01-01

    The Hanford Immobilized Low-Activity Tank Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-level fraction of waste presently contained in Hanford Site tanks. The tank waste is the by-product of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste has been stored in underground single and double-shell tanks. The tank waste is to be retrieved, separated into low and high-activity fractions, and then immobilized by private vendors. The US Department of Energy (DOE) will receive the vitrified waste from private vendors and plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at Hanford until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to issue a Disposal Authorization Statement that would allow the modification of the four existing concrete disposal vaults to provide better access for emplacement of the immobilized low-activity waste (ILAW) containers; filling of the modified vaults with the approximately 5,000 ILAW containers and filler material with the intent to dispose of the containers; construction of the first set of next-generation disposal facilities. The performance assessment activity will continue beyond this assessment. The activity will collect additional data on the geotechnical features of the disposal sites, the disposal facility design and construction, and the long-term performance of the waste. Better estimates of long-term performance will be produced and reviewed on a regular basis. Performance assessments supporting closure of filled facilities will be issued seeking approval of those actions necessary to conclude active disposal facility operations. This report also analyzes the long-term performance of the currently planned disposal system as a basis

  18. Hanford immobilized low-activity tank waste performance assessment

    Energy Technology Data Exchange (ETDEWEB)

    Mann, F.M.

    1998-03-26

    The Hanford Immobilized Low-Activity Tank Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-level fraction of waste presently contained in Hanford Site tanks. The tank waste is the by-product of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste has been stored in underground single and double-shell tanks. The tank waste is to be retrieved, separated into low and high-activity fractions, and then immobilized by private vendors. The US Department of Energy (DOE) will receive the vitrified waste from private vendors and plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at Hanford until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to issue a Disposal Authorization Statement that would allow the modification of the four existing concrete disposal vaults to provide better access for emplacement of the immobilized low-activity waste (ILAW) containers; filling of the modified vaults with the approximately 5,000 ILAW containers and filler material with the intent to dispose of the containers; construction of the first set of next-generation disposal facilities. The performance assessment activity will continue beyond this assessment. The activity will collect additional data on the geotechnical features of the disposal sites, the disposal facility design and construction, and the long-term performance of the waste. Better estimates of long-term performance will be produced and reviewed on a regular basis. Performance assessments supporting closure of filled facilities will be issued seeking approval of those actions necessary to conclude active disposal facility operations. This report also analyzes the long-term performance of the currently planned disposal system as a basis

  19. SYNROC process. A geochemical approach to nuclear waste immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Ringwood, A E; Kesson, S E; Ware, N G; Hibberson, W O; Major, A [Australian National Univ., Canberra. Research School of Earth Sciences

    1979-08-01

    The SYNROC process is proposed to immobilize high-level wastes as dilute solid solutions in the constituent minerals of a synthetic rock formed from a mixture of oxides. New modification of the SYNROC was developed. Experiments showed that the entire spectra of high-level waste elements can be incorporated in the crystal lattices of Ba-hollandite, perovskite and zirconolite. This titanate assemblage has been proved to be exceptionally resistant to hydrothermal leaching, and in this respect, amongst others, it is demonstrably superior to alternative ceramic waste forms and to borosilicate glasses. The relative stability of various waste forms was compared in hydrothermal leaching experiments using both pure water and 10 w/o NaCl solution. Borosilicate glasses were almost completely decomposed and disintegrated after only 24 hours at 350 deg C and 1000 bars, and the extensive loss of hazardous high-level waste elements occurred. The phase pollucite in ceramic waste forms began to decompose at 400 deg C. The hollandite-perovskite-zirconolite SYNROC assemblage was proved to be exceptionally resistant to leaching, surviving invariably the extreme conditions up to 900 deg C and 5000 bars. Geochemical studies of the naturally-occurring minerals containing radwaste elements are relevant to the problem of radiation damage to SYNROC phases. These imply that the 2-particle flux in SYNROC is unlikely to be enough to impair the ability to immobilize radwaste for the required period. The production of SYNROC is explained in detail. The SYNROC phases have the structures analogous to the natural minerals which have survived a variety of geological conditions for millions of years while retaining certain high-level waste elements in their crystal lattices.

  20. Performance Enhancements to the Hanford Waste Treatment and Immobilization Plant Low-Activity Waste Vitrification System

    International Nuclear Information System (INIS)

    Hamel, W. F.; Gerdes, K.; Holton, L. K.; Pegg, I.L.; Bowan, B.W.

    2006-01-01

    The U.S Department of Energy Office of River Protection (DOE-ORP) is constructing a Waste Treatment and Immobilization Plant (WTP) for the treatment and vitrification of underground tank wastes stored at the Hanford Site in Washington State. The WTP comprises four major facilities: a pretreatment facility to separate the tank waste into high level waste (HLW) and low-activity waste (LAW) process streams, a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction, and an analytical laboratory to support the operations of all four treatment facilities. DOE has established strategic objectives to optimize the performance of the WTP facilities and the LAW and HLW waste forms to reduce the overall schedule and cost for treatment and vitrification of the Hanford tank wastes. This strategy has been implemented by establishing performance expectations in the WTP contract for the facilities and waste forms. In addition, DOE, as owner-operator of the WTP facilities, continues to evaluate 1) the design, to determine the potential for performance above the requirements specified in the WTP contract; and 2) improvements in production of the LAW and HLW waste forms. This paper reports recent progress directed at improving production of the LAW waste form. DOE's initial assessment, which is based on the work reported in this paper, is that the treatment rate of the WTP LAW vitrification facility can be increased by a factor of 2 to 4 with a combination of revised glass formulations, modest increases in melter glass operating temperatures, and a second-generation LAW melter with a larger surface area. Implementing these improvements in the LAW waste immobilization capability can benefit the LAW treatment mission by reducing the cost of waste treatment. (authors)

  1. Immobilization of radioactive waste in cement based matrices

    International Nuclear Information System (INIS)

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

    1985-06-01

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

  2. Study of immobilization of waste from treatment of acid waters of a uranium mining facility; Estudo de imobilização de resíduo proveniente de tratamento de águas ácidas de uma instalação de mineração de urânio

    Energy Technology Data Exchange (ETDEWEB)

    Goda, R.T.; Oliveira, A.P. de; Silva, N.C. da; Villegas, R.A.S., E-mail: ricardogoda@gmail.com [Comissão Nacional de Energia Nuclear (LAPOC/CNEN), Pocos de Caldas, MG (Brazil). Laboratorio de Pocos de Caldas; Ferreira, A.M. [Universidade Federal de Alfenas (ICT/UNIFAL), Poços de Caldas, MG (Brazil). Instituto de Ciência e Tecnologia

    2017-07-01

    This study aimed to produce scientific and technical knowledge aiming at the development of techniques to immobilize the waste generated in the treatment of acid waters in the UTM-INB Caldas uranium mining and processing facility using Portland cement. This residue (calcium diuranate - DUCA) contains uranium compounds and metal hydroxides in a matrix of calcium sulfate. It is observed that this material, in contact with the lake of acid waters of the mine's own pit, undergoes resolubilization and, therefore, changes the quality of the acidic water contained therein, changing the treatment parameters. For the study of immobilization of this residue, the mass of water contained in both the residue deposited in the pit of the mine and in the pulp resulting from the treatment of the acid waters was determined. In addition, different DUCA / CEMENT / WATER ratios were used for immobilization and subsequent mechanical strength and leaching tests. The results showed that in the immobilized samples with 50% cement mass condition, no uranium was detected in the leaching tests, and the mechanical strength at compression was 9.4 MPa, which indicates that more studies are needed, but indicate a good capacity to immobilize uranium in cement.

  3. Talc-silicon glass-ceramic waste forms for immobilization of high- level calcined waste

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1993-06-01

    Talc-silicon glass-ceramic waste forms are being evaluated as candidates for immobilization of the high level calcined waste stored onsite at the Idaho Chemical Processing Plant. These glass-ceramic waste forms were prepared by hot isostatically pressing a mixture of simulated nonradioactive high level calcined waste, talc, silicon and aluminum metal additives. The waste forms were characterized for density, chemical durability, and glass and crystalline phase compositions. The results indicate improved density and chemical durability as the silicon content is increased

  4. Immobilization of radioactive wastes in glasses and ceramics

    International Nuclear Information System (INIS)

    Zanotto, E.D.

    1983-01-01

    A large amount of radioactive liquid wastes arises from the reprocessing of spent nuclear fuels to recover uranium and plutonium. Immobilization of such wastes in solid form and disposal of the solidified wastes in safe places, to prevent contamination of the human environment, are topics of considerable interest for both the scientific community and the public in general. The great majority of materials candidate for the encapsulation of radioactive wastes are inorganic non-metalic, such as glasses, glass-ceramics, special cements, calcined ceramics and few more. Among these materials, certain glasses have received special attention, and are being studied for over twenty years. It is estimated that about US$2 billion have already been spent in these studies. The disposal (long term storage) of these solid wastes may be possible in deep geological formations, salt mines, the ocean bed, by evacuation to the outer space, etc. A brief review on the several options avaiable for encapsulation and disposal of high level radioactive liquid wastes is presented, along with the relative merits and disadvantages of the candidate materials for encapsulation. A few suggestions for the solution of the Brazilian problem are advanced. (Author) [pt

  5. Application of the iron-enriched basalt waste form for immobilizing commercial transuranic waste

    International Nuclear Information System (INIS)

    Owen, D.E.

    1981-08-01

    The principal sources of commercial transuranic (TRU) waste in the United States are identified. The physical and chemical nature of the wastes from these sources are discussed. The fabrication technique and properties of iron-enriched basalt, a rock-like waste form developed for immobilizing defense TRU wastes, are discussed. The application of iron-enriched basalt to commercial TRU wastes is discussed. Review of commercial TRU wastes from mixed-oxide fuel fabrication, light water reactor fuel reprocessing, and miscellaneous medical, research, and industrial sources, indicates that iron-enriched basalt is suitable for most types of commercial TRU wastes. Noncombustible TRU wastes are dissolved in the high temperature, oxidizing iron-enriched basalt melt. Combustible TRU wastes are immobilized in iron-enriched basalt by incinerating the wastes and adding the TRU-bearing ash to the melt. Casting and controlled cooling of the melt produces a devitrified, rock-like iron-enriched basalt monolith. Recommendations are given for testing the applicability of iron-enriched basalt to commercial TRU wastes

  6. Immobilization of radioactive waste in cement-based matrices

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  7. Data Packages for the Hanford Immobilized Low-Activity Tank Waste Performance Assessment: 2001 Version

    International Nuclear Information System (INIS)

    MANN, F.M.

    2000-01-01

    Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided

  8. Technologies for in situ immobilization and isolation of radioactive wastes at disposal and contaminated sites

    International Nuclear Information System (INIS)

    1997-11-01

    This report describes technologies that have been developed worldwide and the experiences applied to both waste disposal and contaminated sites. The term immobilization covers both solidification and embedding of wastes

  9. Maintenance Plan for the Hanford Immobilized Low-Activity Tank Waste Performance Assessment

    International Nuclear Information System (INIS)

    MANN, F.M.

    2000-01-01

    The plan for maintaining the Hanford Immobilized Low-Activity Tank Waste Performance Assessment (PA) is described. The plan includes expected work on PA reviews and revisions, waste reports, monitoring, other operational activities, etc

  10. Foaming and Antifoaming in Radioactive Waste Pretreatment and Immobilization

    International Nuclear Information System (INIS)

    Wasan, Darsh T.

    2002-01-01

    Radioactive waste treatment processes usually involve concentration of radionuclides before waste can be immobilized by storing it in stable solid form. Foaming is observed at various stages of waste processing like sludge chemical processing and melter operations. Hence, the objective of this research was to study the mechanisms that produce foaming during nuclear waste treatment, to identify key parameters which aggravate foaming, and to identify effective ways to eliminate or mitigate foaming. Experimental and theoretical investigations of the surface phenomenon, suspension rheology, and bubble generation and interactions that lead to the formation of foam during waste processing were pursued under this EMSP project. Advanced experimental techniques including a novel capillary force balance in conjunction with the combined differential and common interferometry were developed to characterize particle-particle interactions at the foam lamella surfaces as well as inside the foam lamella. Laboratory tests were conducted using a non-radioactive simulant slurry containing high levels of noble metals and mercury similar to the High-Level Waste. We concluded that foaminess of the simulant sludge was due to the presence of colloidal particles such as aluminum, iron, and manganese. We have established the two major mechanisms of formation and stabilization of foams containing such colloidal particles: (1) structural and depletion forces; and (2) steric stabilization due to the adsorbed particles at the surfaces of the foam lamella. Based on this mechanistic understanding of foam generation and stability, an improved antifoam agent was developed by us, since commercial antifoam agents were found to be ineffective in the aggressive physical and chemical environment present in the sludge processing. The improved antifoamer was subsequently tested in a pilot plant at the Savannah River Site (SRS) and was found to be effective. Also, in the SRTC experiment, the irradiated

  11. A Joule-Heated Melter Technology For The Treatment And Immobilization Of Low-Activity Waste

    International Nuclear Information System (INIS)

    Kelly, S.E.

    2011-01-01

    This report is one of four reports written to provide background information regarding immobilization technologies remaining under consideration for supplemental immobilization of Hanford's low-activity waste. This paper provides the reader a general understanding of joule-heated ceramic lined melters and their application to Hanford's low-activity waste.

  12. A JOULE-HEATED MELTER TECHNOLOGY FOR THE TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE

    Energy Technology Data Exchange (ETDEWEB)

    KELLY SE

    2011-04-07

    This report is one of four reports written to provide background information regarding immobilization technologies remaining under consideration for supplemental immobilization of Hanford's low-activity waste. This paper provides the reader a general understanding of joule-heated ceramic lined melters and their application to Hanford's low-activity waste.

  13. Borosilicate glass as a matrix for the immobilization of Savannah River Plant waste

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Wicks, G.G.; Bibler, N.E.

    1982-01-01

    The reference waste form for immobilization of Savannah River Plant (SRP) waste is borosilicate glass. In the reference process, waste is mixed with glass-forming chemicals and melted in a Joule-heated ceramic melter at 1150 0 C. Waste glass made with actual or simulated waste on a small scale and glass made with simulated waste on a large scale confirm that the current reference process and glass-former composition are able to accommodate all SRP waste compositions and can produce a glass with: high waste loading; low leach rates; good thermal stability; high resistance to radiation effects; and good impact resistance. Borosilicate glass has been studied as a matrix for the immobilization of SRP waste since 1974. This paper reviews the results of extensive characterization and performance testing of the glass product. These results show that borosilicate glass is a very suitable matrix for the immobilization of SRP waste. 18 references, 3 figures, 10 tables

  14. 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.

  15. Supplemental Immobilization Cast Stone Technology Development and Waste Form Qualification Testing Plan

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Serne, R. Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pierce, Eric M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chung, Chul-Woo [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Swanberg, David J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-05-31

    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). The pretreatment facility will have the capacity to separate all of the tank wastes into the HLW and LAW fractions, and the HLW Vitrification Facility will have the capacity to vitrify all of the HLW. However, a second immobilization facility will be needed for the expected volume of LAW requiring immobilization. A number of alternatives, including Cast Stone—a cementitious waste form—are being considered to provide the additional LAW immobilization capacity.

  16. Hanford Tank Waste Treatment and Immobilization Plant (WTP) Waste Feed Qualification Program Development Approach - 13114

    Energy Technology Data Exchange (ETDEWEB)

    Markillie, Jeffrey R.; Arakali, Aruna V.; Benson, Peter A.; Halverson, Thomas G. [Hanford Tank Waste Treatment and Immobilization Plant Project, Richland, WA 99354 (United States); Adamson, Duane J.; Herman, Connie C.; Peeler, David K. [Savannah River National Laboratory, Aiken, SC 29808 (United States)

    2013-07-01

    The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is a nuclear waste treatment facility being designed and constructed for the U.S. Department of Energy by Bechtel National, Inc. and subcontractor URS Corporation (under contract DE-AC27-01RV14136 [1]) to process and vitrify radioactive waste that is currently stored in underground tanks at the Hanford Site. A wide range of planning is in progress to prepare for safe start-up, commissioning, and operation. The waste feed qualification program is being developed to protect the WTP design, safety basis, and technical basis by assuring acceptance requirements can be met before the transfer of waste. The WTP Project has partnered with Savannah River National Laboratory to develop the waste feed qualification program. The results of waste feed qualification activities will be implemented using a batch processing methodology, and will establish an acceptable range of operator controllable parameters needed to treat the staged waste. Waste feed qualification program development is being implemented in three separate phases. Phase 1 required identification of analytical methods and gaps. This activity has been completed, and provides the foundation for a technically defensible approach for waste feed qualification. Phase 2 of the program development is in progress. The activities in this phase include the closure of analytical methodology gaps identified during Phase 1, design and fabrication of laboratory-scale test apparatus, and determination of the waste feed qualification sample volume. Phase 3 will demonstrate waste feed qualification testing in support of Cold Commissioning. (authors)

  17. Immobilization of calcium sulfate contained in demolition waste

    International Nuclear Information System (INIS)

    Ambroise, J.; Pera, J.

    2008-01-01

    This paper presents the results of a laboratory study undertaken to examine the treatment of demolition waste containing calcium sulfate by means of calcium sulfoaluminate clinker (CSA). The quantity of CSA necessary to entirely consume calcium sulfate was determined. Using infrared spectrometry analysis and X-ray diffraction, it was shown that calcium sulfate was entirely consumed when the ratio between CSA and calcium sulfate was 4. Standard sand was polluted by 4% calcium sulfate. Two solutions were investigated: ·either global treatment of sand by CSA, ·or immobilization of calcium sulfate by CSA, followed by the introduction of this milled mixture in standard sand. Regardless of the type of treatment, swelling was almost stabilized after 28 days of immersion in water

  18. Investigation of waste form materials suitable for immobilizing actinide elements in high-level waste

    International Nuclear Information System (INIS)

    Hayakawa, Issei; Kamizono, Hiroshi

    1992-07-01

    The microstructure of waste form materials suitable for immobilizing actinide elements can be classified into the following two categories. (1) Actinide elements are immobilized in an crystallized matrix after the formation of solid solution or compounds. (2) Actinide elements are immobilized in a durable material by encapsulation. Based on crystal chemistry, durability data, phase diagrams, compositions of natural minerals, eleven oxide compounds and one non-oxide compound are pointed out to be new candidates included in category (1). The other survey on material compositions, manufacturing conditions and feasibility shows that SiC, glassy carbon, ZrO 2 , Ti-O-Si-C ceramics are preferable matrix materials included in category (2). Polymers and fine powders are suitable as starting materials for the encapsulation of actinide elements because of their excellent sinterability. (author) 50 refs

  19. Immobilized high-level waste interim storage alternatives generation and analysis and decision report

    International Nuclear Information System (INIS)

    CALMUS, R.B.

    1999-01-01

    This report presents a study of alternative system architectures to provide onsite interim storage for the immobilized high-level waste produced by the Tank Waste Remediation System (TWRS) privatization vendor. It examines the contract and program changes that have occurred and evaluates their impacts on the baseline immobilized high-level waste (IHLW) interim storage strategy. In addition, this report documents the recommended initial interim storage architecture and implementation path forward

  20. Immobilization of defense high-level waste: an assessment of technological strategies and potential regulatory goals. Volume I

    International Nuclear Information System (INIS)

    1979-06-01

    An investigation was made of the high-level radioactive waste immobilization technology programs in the U.S. and Europe, and of the associated regulatory programs and waste management perspectives in the countries studied. Purpose was to assess the ability of those programs to satisfy DOE waste management needs and U.S. regulatory requirements. This volume includes: introduction, immobilization strategies in the context of waste isolation program needs, high-level waste management as an integrated system, regulatory goals, engineered-barrier characteristics, barrier technology, high-level waste disposal programs, analysis of HLW immobilization technology in the context of policy and regulatory requirements, and waste immobilization program option

  1. Supplemental Immobilization of Hanford Low-Activity Waste: Cast Stone Screening Tests

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Piepel, Gregory F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lindberg, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Heasler, Patrick G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mercier, Theresa M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Russell, Renee L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daniel, William E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Eibling, Russell E. [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); Reigel, Marissa M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Swanberg, David J. [Washington River Protection Solutions (WRPS), Aiken, SC (United States)

    2013-09-30

    More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in southeastern Washington State. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the wastes and immobilize them in a glass waste form. The WTP includes a pretreatment facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the nonradioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at an offsite federal repository. At least a portion (~35%) of the LAW will be converted to glass in the LAW vitrification facility and will be disposed of onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize the wastes destined for each facility. However, a second LAW immobilization facility will be needed for the expected volume of LAW requiring immobilization. A cementitious waste form known as Cast Stone is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with the waste acceptance criteria for the disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. Further, the waste form must be tested 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 risk assessment and performance assessment (PA) analyses of the long-term environmental impact of the waste disposal in the IDF

  2. Immobilization of high-level wastes into sintered glass: 2

    International Nuclear Information System (INIS)

    Bevilacqua, A.M.; Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

    1987-01-01

    High level radioactive wastes are immobilized into borosilicate glasses. Experiences with the variety VG 98/12 (SiO 2 , TiO 2 , Al 2 O 3 , B 2 O 3 , MgO, CaO, Na 2 O) are described. The pressing was performed in a matrix of 12.7 mm diameter, the walls of which were lubricated with sterotex dissolved in Cl 4 C. The sintering was made in an horizontal electric furnace in air atmosphere at temperatures between 500 and 600 deg C. It was observed that the maximum density occurs at 605 deg C. Comparing both the hot and the cold pressing process, it is concluded that: 1) In spite of requiring more complex equipment the hot pressing process has the advantage that lower pressures are applied, with the consequent obtainment of waste blocks with larger diameters, and 2) it is advisable that pressing process should be performed in the definitive can. (M.E.L.) [es

  3. 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.

  4. Waste immobilization demonstration program for the Hanford Site's Mixed Waste Facility

    International Nuclear Information System (INIS)

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

    1994-05-01

    This paper presents an overview of the Waste Receiving and Processing facility, Module 2A> waste immobilization demonstration program, focusing on the cooperation between Hanford Site, commercial, and international participants. Important highlights of the development and demonstration activities is discussed from the standpoint of findings that have had significant from the standpoint of findings that have had significant impact on the evolution of the facility design. A brief description of the future direction of the program is presented, with emphasis on the key aspects of the technologies that call for further detailed investigation

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

    International Nuclear Information System (INIS)

    Oversby, V.M.; van Konynenburg, R.A.; Glassley, W.E.; Curtis, P.G.

    1993-11-01

    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

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

    International Nuclear Information System (INIS)

    Voss, J.W.

    1979-01-01

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

  7. Exposure Scenarios and Unit Dose Factors for the Hanford Immobilized Low Activity Tank Waste Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    RITTMANN, P.D.

    1999-12-29

    Exposure scenarios are defined to identify potential pathways and combinations of pathways that could lead to radiation exposure from immobilized tank waste. Appropriate data and models are selected to permit calculation of dose factors for each exposure

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

    International Nuclear Information System (INIS)

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

    1987-09-01

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

  9. Design requirements document for project W-520, immobilized low-activity waste disposal

    International Nuclear Information System (INIS)

    Ashworth, S.C.

    1998-01-01

    This design requirements document (DRD) identifies the functions that must be performed to accept, handle, and dispose of the immobilized low-activity waste (ILAW) produced by the Tank Waste Remediation System (TWRS) private treatment contractors and close the facility. It identifies the requirements that are associated with those functions and that must be met. The functional and performance requirements in this document provide the basis for the conceptual design of the Tank Waste Remediation System Immobilized Low-Activity Waste disposal facility project (W-520) and provides traceability from the program-level requirements to the project design activity

  10. Design requirements document for project W-520, immobilized low-activity waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Ashworth, S.C.

    1998-08-06

    This design requirements document (DRD) identifies the functions that must be performed to accept, handle, and dispose of the immobilized low-activity waste (ILAW) produced by the Tank Waste Remediation System (TWRS) private treatment contractors and close the facility. It identifies the requirements that are associated with those functions and that must be met. The functional and performance requirements in this document provide the basis for the conceptual design of the Tank Waste Remediation System Immobilized Low-Activity Waste disposal facility project (W-520) and provides traceability from the program-level requirements to the project design activity.

  11. Waste incineration and immobilization for nuclear facilities, April--September 1977

    International Nuclear Information System (INIS)

    Johnson, A.J.; Fong, L.Q.

    1978-01-01

    Fluidized bed incineration and waste immobilization processes are being developed to process the types of waste expected from nuclear facilities. An air classification system has been developed to separate tramp metal from shredded combustible solid waste prior to the waste being fed to a fluidized-bed pilot-plant incinerator. Used organic ion exchange resin with up to 55 percent water has been effectively burned in the fluidized bed incinerator. Various methods of feeding waste into the incinerator were investigated as alternatives to the present compression screw; an extrusion ram was found to suffer extensive damage from hard particles in tested waste. A bench-scale continuous waste immobilization process has been operated and has produced glass from incinerator residue and other types of waste materials

  12. Immobilization of Radioactive Rare Earth oxide Waste by Solid Phase Sintering

    International Nuclear Information System (INIS)

    Ahn, Byung Gil; Park, Hwan Seo; Kim, Hwan Young; Lee, Han Soo; Kim, In Tae

    2010-01-01

    In the pyroprocessing of spent nuclear fuels, LiCl-KCl waste salt containing radioactive rare earth chlorides are generated. The radioactive rare earth oxides are recovered by co-oxidative precipitation of rare earth elements. The powder phase of rare earth oxide waste must be immobilized to produce a monolithic wasteform suitable for storage and ultimate disposal. The immobilization of these waste developed in this study involves a solid state sintering of the waste with host borosilicate glass and zinc titanate based ceramic matrix (ZIT). And the rare-earth monazite which synthesised by reaction of ammonium di-hydrogen phosphate with the rare earth oxides waste, were immobilized with the borosilicate glass. It is shown that the developed ZIT ceramic wasteform is highly resistant the leaching process, high density and thermal conductivity.

  13. UK program: glasses and ceramics for immobilization of radioactive wastes for disposal

    International Nuclear Information System (INIS)

    Johnson, K.D.B.

    1979-01-01

    The UK Research Program on Radioactive Waste Management includes the development of processes for the conversion of high-level-liquid-reprocessing wastes from thermal and fast reactors to borosilicate glasses. The properties of these glasses and their behavior under storage and disposal conditions have been examined. Methods for immobilizing activity from other wastes by conversion to glass or ceramic forms are described. The UK philosophy of final solutions to waste management and disposal is presented

  14. TWRS retrieval and storage mission. Immobilized low-activity waste disposal plan

    International Nuclear Information System (INIS)

    Shade, J.W.

    1998-01-01

    The TWRS mission is to store, treat, and immobilize highly radioactive Hanford waste (current and future tank waste and the encapsulated cesium and strontium) in a safe, environmentally sound, and cost-effective manner (TWRS JMN Justification for mission need). The mission includes retrieval, pretreatment, immobilization, interim storage and disposal, and tank closure. As part of this mission, DOE has established the TWRS Office to manage all Hanford Site tank waste activities. The TWRS program has identified the need to store, treat, immobilize, and dispose of the highly radioactive Hanford Site tank waste and encapsulated cesium and strontium materials in an environmentally sound, safe, and cost-effective manner. To support environmental remediation and restoration at the Hanford Site a two-phase approach to using private contractors to treat and immobilize the low-activity and high-level waste currently stored in underground tanks is planned. The request for proposals (RFP) for the first phase of waste treatment and immobilization was issued in February 1996 (Wagoner 1996) and initial contracts for two private contractor teams led by British Nuclear Fuels Ltd. and Lockheed-Martin Advanced Environmental Services were signed in September 1996. Phase 1 is a proof-of-concept and commercial demonstration effort to demonstrate the technical and business feasibility of using private facilities to treat Hanford Site waste, maintain radiological, nuclear, process, and occupational safety; and maintain environmental protection and compliance while reducing lifecycle costs and waste treatment times. Phase 1 production of ILAW is planned to begin in June 2002 and could treat up to about 13 percent of the waste. Phase 1 production is expected to be completed in 2007 for minimum order quantities or 2011 for maximum order quantities. Phase 2 is a full-scale production effort that will begin after Phase 1 and treat and immobilize most of the waste. Phase 2 production is

  15. US program for the immobilization of high-level nuclear wastes

    International Nuclear Information System (INIS)

    Crandall, J.L.

    1979-01-01

    A program has been developed for long-term management of high-level nuclear waste. The Savannah River Operations Office of the US Department of Energy is acting as the lead office for this program with technical advice from the E.I. du Pont de Nemours and Company. The purpose of the long-term program is to immobilize the DOE high-level waste in forms that act as highly efficient barriers against radionuclide release to the disposal site and to provide technology for similar treatment of commercial high-level waste in case reprocessing of commercial nuclear fuels is ever resumed. Descriptions of existing DOE and commercial wastes, program strategy, program expenditures, development of waste forms, evaluation and selection of waste forms, regulatory aspects of waste form selection, project schedules, and cost estimates for immobilization facilities are discussed

  16. The Defense Waste Processing Facility: an innovative process for high-level waste immobilization

    International Nuclear Information System (INIS)

    Cowan, S.P.

    1985-01-01

    The Defense Waste Processing Facility (DWPF), under construction at the Department of Energy's Savannah River Plant (SRP), will process defense high-level radioactive waste so that it can be disposed of safely. The DWPF will immobilize the high activity fraction of the waste in borosilicate glass cast in stainless steel canisters which can be handled, stored, transported and disposed of in a geologic repository. The low-activity fraction of the waste, which represents about 90% of the high-level waste HLW volume, will be decontaminated and disposed of on the SRP site. After decontamination the canister will be welded shut by an upset resistance welding technique. In this process a slightly oversized plug is pressed into the canister opening. At the same time a large current is passed through the canister and plug. The higher resistance of the canister/plug interface causes the heat which welds the plug in place. This process provides a high quality, reliable weld by a process easily operated remotely

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

    International Nuclear Information System (INIS)

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

    1994-03-01

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

  18. Radiation pretreatment of cellulosic wastes and immobilization of cells producing cellulase for their conversion to glucose

    International Nuclear Information System (INIS)

    Kumakura, Minoru; Kaetsu, Isao

    1988-01-01

    Radiation pretreatment of cellulosic wastes such as saw dust and chaff was studied by using electron beam accelerator, in which irradiation effect was increased by increasing irradiation dose and dose rate, by after heating irradiated materials at 100∼140deg C, and by irradiation in the addition of alkaline solution. Trichoderma reesei cells producing cellulase were immobilized by using fibrous porous carrier obtained from radiation polymerization. The filter paper, cellobiose, and CMC activities in the immobilized growing cells were higher than those in free cells. The activity in the immobilized cells obtained with hydrophobic carrier was higher than that obtained with hydrophilic one. Durability of the immobilized cells was examined by repeated batch culture. It was found that the enzyme solution produced in the culture of the immobilized cells can hydrolyze effectively saw dust pretreated by radiation. (author)

  19. Technologies for recovery of transuranics and immobilization of non-high-level wastes

    International Nuclear Information System (INIS)

    Richardson, G.L.

    1976-06-01

    This paper supplements the preceding Symposium paper on ''Treatment Technologies for Non-High-Level Wastes (U.S.A.)'' by C. R. Cooley and D. E. Clark (HEDL-SA-851), and covers the additional treatment technologies in use and under development for recovering transuranics and immobilizing non-high-level wastes for transportation and storage. Methods used for nondestructive assay (NDA) of TRU elements in non-high-level wastes are also discussed briefly

  20. Test procedures for polyester immobilized salt-containing surrogate mixed wastes

    International Nuclear Information System (INIS)

    Biyani, R.K.; Hendrickson, D.W.

    1997-01-01

    These test procedures are written to meet the procedural needs of the Test Plan for immobilization of salt containing surrogate mixed waste using polymer resins, HNF-SD-RE-TP-026 and to ensure adequacy of conduct and collection of samples and data. This testing will demonstrate the use of four different polyester vinyl ester resins in the solidification of surrogate liquid and dry wastes, similar to some mixed wastes generated by DOE operations

  1. Preliminary evaluation of alternative forms for immobilization of Hanford high-level defense wastes

    International Nuclear Information System (INIS)

    Schulz, W.W.; Beary, M.M.; Gallagher, S.A.; Higley, B.A.; Johnston, R.G.; Jungfleisch, F.M.; Kupfer, M.J.; Palmer, R.A.; Watrous, R.A.; Wolf, G.A.

    1980-09-01

    A preliminary evaluation of solid waste forms for immobilization of Hanford high-level radioactive defense wastes is presented. Nineteen different waste forms were evaluated and compared to determine their applicability and suitability for immobilization of Hanford salt cake, sludge, and residual liquid. This assessment was structured to address waste forms/processes for several different leave-retrieve long-term Hanford waste management alternatives which give rise to four different generic fractions: (1) sludge plus long-lived radionuclide concentrate from salt cake and residual liquid; (2) blended wastes (salt cake plus sludge plus residual liquid); (3) residual liquid; and (4) radionuclide concentrate from residual liquid. Waste forms were evaluated and ranked on the basis of weighted ratings of seven waste form and seven process characteristics. Borosilicate Glass waste forms, as marbles or monoliths, rank among the first three choices for fixation of all Hanford high-level wastes (HLW). Supergrout Concrete (akin to Oak Ridge National Laboratory Hydrofracture Process concrete) and Bitumen, low-temperature waste forms, rate high for bulk disposal immobilization of high-sodium blended wastes and residual liquid. Certain multi-barrier (e.g., Coated Ceramic) and ceramic (SYNROC Ceramic, Tailored Ceramics, and Supercalcine Ceramic) waste forms, along with Borosilicate Glass, are rated as the most satisfactory forms in which to incorporate sludges and associated radionuclide concentrates. The Sol-Gel process appears superior to other processes for manufacture of a generic ceramic waste form for fixation of Hanford sludge. Appropriate recommendations for further research and development work on top ranking waste forms are made

  2. Ethanol production from concentrated food waste hydrolysates with yeast cells immobilized on corn stalk

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Shoubao [Huainan Normal Univ., Anhui (China). School of Life Science; Chen, Xiangsong; Wu, Jingyong; Wang, Pingchao [Chinese Academy of Sciences, Hefei (China). Key Lab. of Ion Beam Bio-engineering of Inst. of Plasma Physics

    2012-05-15

    The aim of the present study was to examine ethanol production from concentrated food waste hydrolysates using whole cells of S. cerevisiae immobilized on corn stalks. In order to improve cell immobilization efficiency, biological modification of the carrier was carried out by cellulase hydrolysis. The results show that proper modification of the carrier with cellulase hydrolysis was suitable for cell immobilization. The mechanism proposed, cellulase hydrolysis, not only increased the immobilized cell concentration, but also disrupted the sleek surface to become rough and porous, which enhanced ethanol production. In batch fermentation with an initial reducing sugar concentration of 202.64 {+-} 1.86 g/l, an optimal ethanol concentration of 87.91 {+-} 1.98 g/l was obtained using a modified corn stalk-immobilized cell system. The ethanol concentration produced by the immobilized cells was 6.9% higher than that produced by the free cells. Ethanol production in the 14th cycle repeated batch fermentation demonstrated the enhanced stability of the immobilized yeast cells. Under continuous fermentation in an immobilized cell reactor, the maximum ethanol concentration of 84.85 g/l, and the highest ethanol yield of 0.43 g/g (of reducing sugar) were achieved at hydraulic retention time (HRT) of 3.10 h, whereas the maximum volumetric ethanol productivity of 43.54 g/l/h was observed at a HRT of 1.55 h. (orig.)

  3. SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. [Savannah River National Laboratory, Aiken, SC (United States); Marra, J. [Savannah River National Laboratory, Aiken, SC (United States)

    2014-10-02

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing.

  4. SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization

    International Nuclear Information System (INIS)

    Amoroso, J.; Marra, J.

    2014-01-01

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear fuel. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing

  5. Data Packages for the Hanford Immobilized Low Activity Tank Waste Performance Assessment 2001 Version [SEC 1 THRU 5

    Energy Technology Data Exchange (ETDEWEB)

    MANN, F.M.

    2000-03-02

    Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided.

  6. Application of eggshell waste for the immobilization of cadmium and lead in a contaminated soil.

    Science.gov (United States)

    Ok, Yong Sik; Lee, Sang Soo; Jeon, Weon-Tai; Oh, Sang-Eun; Usman, Adel R A; Moon, Deok Hyun

    2011-01-01

    Liming materials have been used to immobilize heavy metals in contaminated soils. However, no studies have evaluated the use of eggshell waste as a source of calcium carbonate (CaCO₃) to immobilize both cadmium (Cd) and lead (Pb) in soils. This study was conducted to evaluate the effectiveness of eggshell waste on the immobilization of Cd and Pb and to determine the metal availability following various single extraction techniques. Incubation experiments were conducted by mixing 0-5% powdered eggshell waste and curing the soil (1,246 mg Pb kg⁻¹ soil and 17 mg Cd kg⁻¹ soil) for 30 days. Five extractants, 0.01 M calcium chloride (CaCl₂), 1 M CaCl₂, 0.1 M hydrochloric acid (HCl), 0.43 M acetic acid (CH₃COOH), and 0.05 M ethylendiaminetetraacetic acid (EDTA), were used to determine the extractability of Cd and Pb following treatments with CaCO₃ and eggshell waste. Generally, the extractability of Cd and Pb in the soils decreased in response to treatments with CaCO₃ and eggshell waste, regardless of extractant. Using CaCl₂ extraction, the lowest Cd concentration was achieved upon both CaCO₃ and eggshell waste treatments, while the lowest Pb concentration was observed using HCl extraction. The highest amount of immobilized Cd and Pb was extracted by CH₃COOH or EDTA in soils treated with CaCO₃ and eggshell waste, indicating that remobilization of Cd and Pb may occur under acidic conditions. Based on the findings obtained, eggshell waste can be used as an alternative to CaCO₃ for the immobilization of heavy metals in soils.

  7. Immobilized low-activity waste interim storage facility, Project W-465 conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Pickett, W.W.

    1997-12-30

    This report outlines the design and Total Estimated Cost to modify the four unused grout vaults for the remote handling and interim storage of immobilized low-activity waste (ILAW). The grout vault facilities in the 200 East Area of the Hanford Site were constructed in the 1980s to support Tank Waste disposal activities. The facilities were to serve project B-714 which was intended to store grouted low-activity waste. The existing 4 unused grout vaults, with modifications for remote handling capability, will provide sufficient capacity for approximately three years of immobilized low activity waste (ILAW) production from the Tank Waste Remediation System-Privatization Vendors (TWRS-PV). These retrofit modifications to the grout vaults will result in an ILAW interim storage facility (Project W465) that will comply with applicable DOE directives, and state and federal regulations.

  8. Immobilized low-activity waste interim storage facility, Project W-465 conceptual design report

    International Nuclear Information System (INIS)

    Pickett, W.W.

    1997-01-01

    This report outlines the design and Total Estimated Cost to modify the four unused grout vaults for the remote handling and interim storage of immobilized low-activity waste (ILAW). The grout vault facilities in the 200 East Area of the Hanford Site were constructed in the 1980s to support Tank Waste disposal activities. The facilities were to serve project B-714 which was intended to store grouted low-activity waste. The existing 4 unused grout vaults, with modifications for remote handling capability, will provide sufficient capacity for approximately three years of immobilized low activity waste (ILAW) production from the Tank Waste Remediation System-Privatization Vendors (TWRS-PV). These retrofit modifications to the grout vaults will result in an ILAW interim storage facility (Project W465) that will comply with applicable DOE directives, and state and federal regulations

  9. Immobilization of industrial waste in cement–bentonite clay matrix

    Indian Academy of Sciences (India)

    Results of a series of experimental tests performed to determine the influence of matrix characteristics on the leaching mechanism of copper aluminum oxychloride immobilized into cement matrices are presented. The objective of this work was to investigate the leaching mechanism of copper as a constituent of copper ...

  10. Assessment of methods for immobilizing reprocessed radioactive waste

    Science.gov (United States)

    Murthy, M. K.; Baranyi, A. D.

    1980-05-01

    Nuclear waste forms presently used for the disposal of high level wastes and other potential waste forms under development were studied. The following waste forms were considered: Borosilicate glass, high silica glass, glassceramics, supercalcine ceramics, synroc ceramics, borosilicate glass beads in a metal matrix, supercalcine and synroc ceramics in a metal matrix and coated ceramics. The best developed wasteform, both in terms of overall product quality and process development, is monolithic borosilicate glass. However, hydrothermal instability is a major concern. Borosilicate glass in metal matrix waste form has better properties than monolithic borosilicate glass waste form. The process was proven on a pilot scale. Hence, it is considered very close to monolithic glass in terms of overall development. The product qualities of the other waste forms are better than borosilicate glass. However, process development for these alternative waste forms is still in a conceptual stage.

  11. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    International Nuclear Information System (INIS)

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-01-01

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE's mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies

  12. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    Energy Technology Data Exchange (ETDEWEB)

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-07-07

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE`s mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies.

  13. Preparation of Metal Immobilized Orange Waste Gel for Arsenic(V Removal From Water

    Directory of Open Access Journals (Sweden)

    Biplob Kumar Biswas

    2014-05-01

    Full Text Available Abstract - The toxicity of arsenic is known to be a risk to aquatic flora and fauna and to human health even in relatively low concentration. In this research an adsorption gel was prepared from agricultural waste material (orange waste through simple chemical modification in the view to remove arsenic (V from water. Orange waste was crushed into small particles and saponified with Ca(OH2 to prepare saponified orange waste, which was further modified by immobilizing gadolinium(III to obtain desired adsorption material (Gd(III-immobilized SOW gel. The effective pH range for arsenic adsorption was found to be 7.5 – 8.5. Adsorption capacity of the gel was evaluated to be 0.45 mol-arsenic (V/kg. Dynamic adsorption of arsenic (V in column-mode was conducted and a dynamic capacity was found to be 0.39 mol/kg. Elution of arsenate was tested after complete saturation of the column packed with gadolinium-immobilized orange waste adsorption gel. A complete elution of arsenate was achieved with the help of 1 M HCl and 28 times pre-concentration factor was attained. This study showed that a cheap and abundant agro-industrial waste material could be successfully employed for the remediation of arsenic pollution in aquatic environment. Keywords: Arsenic; Orange waste; Gadolinium(III; Adsorption; Elution.

  14. Feasibility of converting lactic acid to ethanol in food waste fermentation by immobilized lactate oxidase

    International Nuclear Information System (INIS)

    Ma, Hong-zhi; Xing, Yi; Yu, Miao; Wang, Qunhui

    2014-01-01

    Highlights: • Residue lactic acid in food waste could be converted to pyruvic acid. • Calcium alginate immobilized the lactate oxidase with high pH and thermal stability. • Immobilized enzyme could convert 70% lactic acid to pyruvic acid. • Ethanol yield could be increased by 20% with lactate oxidase added. - Abstract: Adoption of lactic acid bacteria (LAB) into ethanol fermentation from food waste can replace the sterilization process. However, LAB inoculation will convert part of the substrate into lactic acid (LA), not ethanol. This study adopted lactate oxidase to convert the produced LA to pyruvate, and then ethanol fermentation was carried out. The immobilization enzyme was utilized, and corresponding optimum conditions were determined. Results showed that calcium alginate could successfully immobilize the enzyme and improve pH and thermal stability. The optimum pH and temperature were 6.2 and 55 °C, respectively. The utilization of immobilized enzyme with catalytic time of 5 h could convert 70% LA to pyruvate, and the addition of enzyme increased the ethanol yield by 20% more than that of the control. The process could be applied in food waste storage and can help in reducing carbon source consumption

  15. Foaming and Antifoaming in Radioactive Waste Pretreatment and Immobilization Processes

    International Nuclear Information System (INIS)

    Wasan, Darsh T.; Nikolov, Alex D.; Lamber, D.P.; Calloway, T. Bond; Stone, M.E.

    2005-01-01

    Savannah River National Laboratory (SRNL) has reported severe foaminess in the bench scale evaporation of the Hanford River Protection - Waste Treatment Plant (RPP-WPT) envelope C waste. Excessive foaming in waste evaporators can cause carryover of radionuclides and non-radioactive waste to the condensate system. The antifoams used at Hanford and tested by SRNL are believed to degrade and become inactive in high pH solutions. Hanford wastes have been known to foam during evaporation causing excessive down time and processing delays

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

    International Nuclear Information System (INIS)

    Glasser, F.P.

    1985-01-01

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

  17. An assessment of methods for immobilizing reprocessed radioactive waste

    International Nuclear Information System (INIS)

    Murthy, M.K.; Baranyi, A.D.

    1980-05-01

    Nuclear waste forms presently used for the disposal of high-level wastes and other potential waste forms under development were studied using information available in the literature and by visits to the laboratories. The following waste forms were considered: Borosilicate glass, high-silica glass, glassceramics, supercalcine ceramics, synroc ceramics, borosilicate glass beads in a metal matrix, supercalcine and synroc ceramics in a metal matrix and coated ceramics. The following conclusions have been reached: To date the best developed wasteform, both in terms of overall product quality and process development, is monolithic borosilicate glass. However, hydrothermal instability is a major concern. Borosilicate glass in metal matrix waste form has better properties than monolithic borosilicate glass waste form. The process has been proven on a pilot scale. Hence, it is considered very close to monolithic glass in terms of overall development. The product qualities of the other waste forms are better than borosilicate glass. However, process development for these alternative waste forms is still in a conceptual stage. The technological basis for processing ceramic waste forms exists in a well developed state. Nevertheless, adaptation of the technology to continuous hot-cell operation, although feasible, has not been demonstrated. In view of the product potential of ceramic waste forms it is felt that their development should be given emphasis at this time. (auth)

  18. Method of immobilizing weapons plutonium to provide a durable, disposable waste product

    Science.gov (United States)

    Ewing, Rodney C.; Lutze, Werner; Weber, William J.

    1996-01-01

    A method of atomic scale fixation and immobilization of plutonium to provide a durable waste product. Plutonium is provided in the form of either PuO.sub.2 or Pu(NO.sub.3).sub.4 and is mixed with and SiO.sub.2. The resulting mixture is cold pressed and then heated under pressure to form (Zr,Pu)SiO.sub.4 as the waste product.

  19. Waste incineration and immobilization for nuclear facilities. Status report, October 1977--March 1978

    International Nuclear Information System (INIS)

    Johnson, A.J.; Burkhardt, S.C.; Ledford, J.A.; Williams, P.M.

    1979-01-01

    Fluidized bed incineration and processes for immobilization of wastes generated at nuclear facilities are undergoing development. After minor piping modifications to eliminate dust collecting points, a pilot plant fluidized bed incinerator run of 225 continuous hours was successfully completed in a demonstration of component reliability. Vitrification of incinerator ash and other wastes is now being accomplished using a pilot scale unit developed as a continuous flow process

  20. Glass science tutorial: Lecture No. 8, introduction cementitious systems for Low-Level Waste immobilization

    International Nuclear Information System (INIS)

    Young, J.F.; Kirkpatrick, R.J.; Mason, T.O.; Brough, A.

    1995-07-01

    This report presents details about cementitious systems for low-level waste immobilization. Topics discussed include: composition and properties of portland cement; hydration properties; microstructure of concrete; pozzolans; slags; zeolites; transport properties; and geological aspects of long-term durability of concrete

  1. 75 FR 81250 - Pulse Jet Mixing at the Waste Treatment and Immobilization Plant

    Science.gov (United States)

    2010-12-27

    ... Immobilization Plant (WTP) in conjunction with the Hanford tank farm waste feed delivery system will operate... imperative requires that the pulse jet mixing and transfer systems relied upon in the WTP design perform reliably and effectively for decades of WTP operations, and that technical issues with the performance of...

  2. Cast Stone Technology For The Treatment And Immobilization Of Low-Activity Waste

    International Nuclear Information System (INIS)

    Minwall, H.J.

    2011-01-01

    Cast stone technology is being evaluated for potential application in the treatment and immobilization of Hanford low-activity waste. The purpose of this document is to provide background information on cast stone technology. The information provided in the report is mainly based on a pre-conceptual design completed in 2003.

  3. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    International Nuclear Information System (INIS)

    Burgard, K.C.

    1998-01-01

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis

  4. CAST STONE TECHNOLOGY FOR THE TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE

    Energy Technology Data Exchange (ETDEWEB)

    MINWALL HJ

    2011-04-08

    Cast stone technology is being evaluated for potential application in the treatment and immobilization of Hanford low-activity waste. The purpose of this document is to provide background information on cast stone technology. The information provided in the report is mainly based on a pre-conceptual design completed in 2003.

  5. Glass science tutorial: Lecture No. 8, introduction cementitious systems for Low-Level Waste immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Young, J.F.; Kirkpatrick, R.J.; Mason, T.O.; Brough, A.

    1995-07-01

    This report presents details about cementitious systems for low-level waste immobilization. Topics discussed include: composition and properties of portland cement; hydration properties; microstructure of concrete; pozzolans; slags; zeolites; transport properties; and geological aspects of long-term durability of concrete.

  6. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-04-09

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  7. The Remote Handled Immobilization Low-Activity Waste Disposal Facility Environmental Permits and Approval Plan

    International Nuclear Information System (INIS)

    DEFFENBAUGH, M.L.

    2000-01-01

    The purpose of this document is to revise Document HNF-SD-ENV-EE-003, ''Permitting Plan for the Immobilized Low-Activity Waste Project, which was submitted on September 4, 1997. That plan accounted for the interim storage and disposal of Immobilized-Low Activity Waste at the existing Grout Treatment Facility Vaults (Project W-465) and within a newly constructed facility (Project W-520). Project W-520 was to have contained a combination of concrete vaults and trenches. This document supersedes that plan because of two subsequent items: (1) A disposal authorization that was received on October 25, 1999, in a U. S. Department of Energy-Headquarters, memorandum, ''Disposal Authorization Statement for the Department of Energy Hanford site Low-Level Waste Disposal facilities'' and (2) ''Breakthrough Initiative Immobilized Low-Activity Waste (ILAW) Disposal Alternative,'' August 1999, from Lucas Incorporated, Richland, Washington. The direction within the U. S. Department of Energy-Headquarters memorandum was given as follows: ''The DOE Radioactive Waste Management Order requires that a Disposal authorization statement be obtained prior to construction of new low-level waste disposal facility. Field elements with the existing low-level waste disposal facilities shall obtain a disposal authorization statement in accordance with the schedule in the complex-wide Low-Level Waste Management Program Plan. The disposal authorization statement shall be issued based on a review of the facility's performance assessment and composite analysis or appropriate CERCLA documentation. The disposal authorization shall specify the limits and conditions on construction, design, operations, and closure of the low-level waste facility based on these reviews. A disposal authorization statement is a part of the required radioactive waste management basis for a disposal facility. Failure to obtain a disposal authorization statement or record of decision shall result in shutdown of an operational

  8. Waste Form Release Data Package for the 2001 Immobilized Low-Activity Waste Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    McGrail, B. Peter; Icenhower, Jonathan P.; Martin, Paul F.; Schaef, Herbert T.; O' Hara, Matthew J.; Rodriguez, Eugenio; Steele, Jackie L.

    2001-02-01

    This data package documents the experimentally derived input data on the representative waste glasses LAWABP1 and HLP-31 that will be used for simulations of the immobilized lowactivity waste disposal system with the Subsurface Transport Over Reactive Multiphases (STORM) code. The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in March of 2001. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali-H ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow and vapor hydration experiments were used for accelerated weathering or aging of the glasses. The majority of the thermodynamic data were extracted from the thermodynamic database package shipped with the geochemical code EQ3/6. However, several secondary reaction products identified from laboratory tests with prototypical LAW glasses were not included in this database, nor are the thermodynamic data available in the open literature. One of these phases, herschelite, was determined to have a potentially significant impact on the release calculations and so a solubility product was estimated using a polymer structure model developed for zeolites. Although this data package is relatively complete, final selection of ILAW glass compositions has not been done by the waste treatment plant contractor. Consequently, revisions to this data package to address new ILAW glass formulations are to be regularly expected.

  9. Immobilization of fission products in phosphate ceramic waste forms

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.

    1997-01-01

    Argonne National Laboratory (ANL) is developing chemically bonded phosphate ceramics (CBPCs) to treat low-level mixed wastes, particularly those containing volatiles and pyrophorics that cannot be treated by conventional thermal processes. This work was begun under ANL''s Laboratory Directed Research and Development funds, followed by further development with support from EM-50''s Mixed Waste Focus Area

  10. Cementitious Composites for Immobilization of Radioactive Waste into Final Wasteform

    International Nuclear Information System (INIS)

    Varlakov, A.P.

    2013-01-01

    Research and development works are important on universal cementation technological processes to achieve maximal conditioning efficiency for various type wastes such as saline liquid radioactive waste (LRW), where the variants of cement composition formulations, modes of cement compounds preparation and types of equipment are minimised. This work presents the results of development of multi-component cement compositions for the complex of technological processes of different types of radioactive waste (RAW) cementation: concentrated saline LRW, concentrated boron-containing saline LRW, LRW with high surface active substances content, with residues, liquid organic radioactive waste, spent ion-exchange resins and filter-perlite powder, ash residues from solid radioactive waste (SRW) combustion, mixed closely packed and large-fragmented SRW. The research has found technological parameters of equipment and cement compositions providing reliable RAW cementation. Continuous and periodic cycle plants were developed for LRW cementation by mixing. Pouring and penetration methods were developed for SRW cementation. Based on compliance with equipment parameters, methods and cement grouts were selected for most effective technological processes of cementation. Formulations of cement compositions were developed to provide reliable preparation of cement compounds with maximal waste loading at required cement compound quality. The complex of technological processes of cementation using multi-component cement compositions allows highly efficient treatment of the wide range of RAW including problematic waste streams and wastes generated in small amounts. Rational reduction of cementation variants significantly increases economical efficiency of immobilisation. (author)

  11. Radiation effects in glass and glass-ceramic waste forms for the immobilization of CANDU UO2 fuel reprocessing waste

    International Nuclear Information System (INIS)

    Tait, J.C.

    1993-05-01

    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 129 I, 85 Kr and 14 C. (author). 104 refs., 9 tabs., 5 figs

  12. Immobilization of zinc from metallurgical waste and water solutions using geopolymerization technology

    Directory of Open Access Journals (Sweden)

    Nikolići I.

    2014-07-01

    Full Text Available Geopolymeraization technology is recognized as a promising method for immobilization of heavy metals by the stabilization or solidification process. This process involves the chemical reaction of alumino-silicate oxides with highly alkaline activator yielding the new material with amorphous or semi-amorphous structure, called geopolymer. Fly ash and blast furnace slag were mainly used as a raw material for geopolymerization process. In this paper we have investigated the possibility of immobilization of Zn from electric arc furnace dust (EAFD through geopolymerization of fly ash and possibility of Zn2+ adsorption from waste waters using fly ash based geopolymers. Efficacy of Zn immobilization from electric arc furnace dust was evaluated by TCLP test while the immobilization of Zn2+ ions from the water solution was evaluated through the removal efficiency. The results have shown that geopolymerization process may successfully be used for immobilization of Zn by stabilization of EAFD and for production of low cost adsorbent for waste water treatment.

  13. Conceptual process for immobilizing defense high level wastes in SYNROC-D

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    It is believed that the immobilization of defense wastes in SYNROC-D possesses important advantages over an alternative process which involves immobilizing the sludges in borosilicate glass. (1) It is possible to immobilize about 3 times the weight of sludge in a given volume of SYNROC-D as compared to borosilicate glass. The costs of fabrications, transport and ultimate geologic storage are correspondingly reduced; (2) the mineral assemblage of SYNROC-D is vastly more stable in the presence of groundwaters than are borosilicate glasses. The long-lived actinide elements, in particular, are immobilized much more securely in SYNROC-D than in glass; and (3) SYNROC-D is composed of thermodynamically compatible phases which possess crystal structures identical to those of natural minerals which are known to have survived in geological environments at elevated pressures and temperatures for periods of 500 to 2000 million years and to have retained radioactive elements quantitatively for these periods despite strong radiation damage. It is this evidence, provided by nature herself, which can demonstrate to the community that the shorter times required for radwaste immobilization under the much less extreme pressure, temperature conditions present in a suitable geological repository can be successfully achieved. Glass, as a waste-form, is intrinsically incapable of providing this assurance

  14. A Glass-Ceramic Waste Forms for the Immobilization of Rare Earth Oxides from the Pyroprocessing Waste salt

    International Nuclear Information System (INIS)

    Ahn, Byung-Gil; Park, Hwan-Seo; Kim, Hwan-Young; Kim, In-Tae

    2008-01-01

    The fission product of rare earth (RE) oxide wastes are generates during the pyroprocess . Borosilicate glass or some ceramic materials such as monazite, apatite or sodium zirconium phosphate (NZP) have been a prospective host matrix through lots of experimental results. Silicate glasses have long been the preferred waste form for the immobilization of HLW. In immobilization of the RE oxides, the developed process on an industrial scale involves their incorporation into a glass matrix, by melting under 1200 ∼ 1300 .deg. C. Instead of the melting process, glass powder sintering is lower temperature (∼ 900 .deg. C) required for the process which implies less demanding conditions for the equipment and a less evaporation of volatile radionuclides. This study reports the behaviors, direct vitrification of RE oxides with glass frit, glass powder sintering of REceramic with glass frit, formation of RE-apatite (or REmonazite) ceramic according to reaction temperature, and the leach resistance of the solidified waste forms

  15. The removal of thermo-tolerant coliform bacteria by immobilized waste stabilization pond algae.

    Science.gov (United States)

    Pearson, H W; Marcon, A E; Melo, H N

    2011-01-01

    This study investigated the potential of laboratory- scale columns of immobilized micro-algae to disinfect effluents using thermo-tolerant coliforms (TTC) as a model system. Cells of a Chlorella species isolated from a waste stabilization pond complex in Northeast Brazil were immobilized in calcium alginate, packed into glass columns and incubated in contact with TTC suspensions for up to 24 hours. Five to six log removals of TTC were achieved in 6 hours and 11 log removals in 12 hours contact time. The results were similar under artificial light and shaded sunlight. However little or no TTC removal occurred in the light in columns of alginate beads without immobilized algae present or when the immobilized algae were incubated in the dark suggesting that the presence of both algae and light were necessary for TTC decay. There was a positive correlation between K(b) values for TTC and increasing pH in the effluent from the immobilized algal columns within the range pH 7.2 and 8.9. The potential of immobilized algal technology for wastewater disinfection may warrant further investigation.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. Electron Beam-Induced Immobilization of Laccase on Porous Supports for Waste Water Treatment Applications

    Directory of Open Access Journals (Sweden)

    Elham Jahangiri

    2014-08-01

    Full Text Available The versatile oxidase enzyme laccase was immobilized on porous supports such as polymer membranes and cryogels with a view of using such biocatalysts in bioreactors aiming at the degradation of environmental pollutants in wastewater. Besides a large surface area for supporting the biocatalyst, the aforementioned porous systems also offer the possibility for simultaneous filtration applications in wastewater treatment. Herein a “green” water-based, initiator-free, and straightforward route to highly reactive membrane and cryogel-based bioreactors is presented, where laccase was immobilized onto the porous polymer supports using a water-based electron beam-initiated grafting reaction. In a second approach, the laccase redox mediators 2,2'-azino-bis(3-ethylbenzothiazoline-6-sulphonic acid (ABTS and syringaldehyde were cross-linked instead of the enzyme via electron irradiation in a frozen aqueous poly(acrylate mixture in a one pot set-up, yielding a mechanical stable macroporous cryogel with interconnected pores ranging from 10 to 50 µm in size. The membranes as well as the cryogels were characterized regarding their morphology, chemical composition, and catalytic activity. The reactivity towards waste- water pollutants was demonstrated by the degradation of the model compound bisphenol A (BPA. Both membrane- and cryogel-immobilized laccase remained highly active after electron beam irradiation. Apparent specific BPA removal rates were higher for cryogel- than for membrane-immobilized and free laccase, whereas membrane-immobilized laccase was more stable with respect to maintenance of enzymatic activity and prevention of enzyme leakage from the carrier than cryogel-immobilized laccase. Cryogel-immobilized redox mediators remained functional in accelerating the laccase-catalyzed BPA degradation, and especially ABTS was found to act more efficiently in immobilized than in freely dissolved state.

  18. Batch Fermentative Biohydrogen Production Process Using Immobilized Anaerobic Sludge from Organic Solid Waste

    Directory of Open Access Journals (Sweden)

    Patrick T. Sekoai

    2016-12-01

    Full Text Available This study examined the potential of organic solid waste for biohydrogen production using immobilized anaerobic sludge. Biohydrogen was produced under batch mode at process conditions of 7.9, 30.3 °C and 90 h for pH, temperature and fermentation time, respectively. A maximum biohydrogen fraction of 48.67%, which corresponded to a biohydrogen yield of 215.39 mL H2/g Total Volatile Solids (TVS, was achieved. Therefore, the utilization of immobilized cells could pave the way for a large-scale biohydrogen production process.

  19. Immobilization of radioactive waste sludge from spent fuel storage pool

    International Nuclear Information System (INIS)

    Pavlovic, R.; Plecas, I.

    1998-01-01

    In the last forty years, in FR Yugoslavia, as result of the research reactors' operation and radionuclides application in medicine, industry and agriculture, radioactive waste materials of the different categories and various levels of specific activities were generated. As a temporary solution, these radioactive waste materials are stored in the two hanger type interim storages for solid waste and some type of liquid waste packed in plastic barrels, and one of three stainless steal underground containers for other types of liquid waste. Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some aspects of immobilisation, conditioning and storage of this sludge are presented in this paper. (author

  20. Immobilization technologies for the management of hazardous industrial waste using granite waste (case study)

    Energy Technology Data Exchange (ETDEWEB)

    Lasheen, Mohamed R.; Ashmawy, Azza M.; Ibrahim, Hanan S.; Moniem, Shimaa M. Abdel [National Research Centre, Giza (Egypt)

    2016-03-15

    Full characterization of granite waste sludge (GWS) was accomplished by X-ray diffraction (XRD) and Xray fluorescence (XRF) for identification of its phase and chemical composition. Different leaching tests were conducted to determine the efficiency of the GWS for metal stabilization in hazardous sludge. The leaching of the metals from stabilized contaminated sludge was decreased as the GWS amount increased. Only 15% of GWS was sufficient for stabilization of all metal ions under investigation. The main reason for metal immobilization was attributed to the aluminosilicates or silicates matrix within the GWS, which can transform the metals in the form of their insoluble hydroxides or absorbed in the stabilized matrix. Also, solidification/stabilization technique was used for remediation of contaminated sludge. Compressive strength test after curing for 28 days was used for measuring the effectiveness of remediation technique; it was found to be 1.88MPa. This indicated that the remediated sludge was well solidified and safe to be used as a raw substance for roadway blocks. Therefore, this huge amount of by-product sludge derived from the granite cutting industry, which has a negative environmental impact due to its disposal, can be utilized as a binder material for solidification/stabilization of hazardous sludge.

  1. Immobilization of simulated radioactive soil waste containing cerium by self-propagating high-temperature synthesis

    Science.gov (United States)

    Mao, Xianhe; Qin, Zhigui; Yuan, Xiaoning; Wang, Chunming; Cai, Xinan; Zhao, Weixia; Zhao, Kang; Yang, Ping; Fan, Xiaoling

    2013-11-01

    A simulated radioactive soil waste containing cerium as an imitator element has been immobilized by a thermite self-propagating high-temperature synthesis (SHS) process. The compositions, structures, and element leaching rates of products with different cerium contents have been characterized. To investigate the influence of iron on the chemical stability of the immobilized products, leaching tests of samples with different iron contents with different leaching solutions were carried out. The results showed that the imitator element cerium mainly forms the crystalline phases CeAl11O18 and Ce2SiO5. The leaching rate of cerium over a period of 28 days was 10-5-10-6 g/(m2 day). Iron in the reactants, the reaction products, and the environment has no significant effect on the chemical stability of the immobilized SHS products.

  2. Immobilization of simulated radioactive soil waste containing cerium by self-propagating high-temperature synthesis

    International Nuclear Information System (INIS)

    Mao, Xianhe; Qin, Zhigui; Yuan, Xiaoning; Wang, Chunming; Cai, Xinan; Zhao, Weixia; Zhao, Kang; Yang, Ping; Fan, Xiaoling

    2013-01-01

    A simulated radioactive soil waste containing cerium as an imitator element has been immobilized by a thermite self-propagating high-temperature synthesis (SHS) process. The compositions, structures, and element leaching rates of products with different cerium contents have been characterized. To investigate the influence of iron on the chemical stability of the immobilized products, leaching tests of samples with different iron contents with different leaching solutions were carried out. The results showed that the imitator element cerium mainly forms the crystalline phases CeAl 11 O 18 and Ce 2 SiO 5 . The leaching rate of cerium over a period of 28 days was 10 −5 –10 −6 g/(m 2 day). Iron in the reactants, the reaction products, and the environment has no significant effect on the chemical stability of the immobilized SHS products

  3. Radioactive waste immobilization in protective ceramic forms by the HIP method at high pressures

    International Nuclear Information System (INIS)

    Sayenko, S.Yu.; Kantsedal, V.P.; Tarasov, R.V.; Starchenko, V.A.; Lyubtsev, R.I.

    1993-01-01

    Intense research activities have been carried out in recent years at the Kharkov Institute of Physics and Technology (KIPT) to develop the method of hot isostatic pressing (HIP) for immobilizing radioactive (primarily, high-level) wastes. With this method, the radioactive material is immobilized in a matrix under the simultaneous action of high pressures (up to 6,000 atm) and appropriate temperatures. The process has 2 variants: (1) radioactive wastes are treated as powders of oxides resulting from calcination during chemical treatment of spent fuel. In this case the radioactive material enters into the crystalline structure of the immobilized matrix or is distributed in the matrix as a homogeneous mixture; (2) protective barrier layers are pressed on spent fuel rods or their pieces as radioactive wastes, by the HIP method (fuel rod encapsulation in a protective form). Based on numerous results from various studies, the authors suggest that various ceramic compositions should be used as protective materials. Here the authors report two trends of their investigations: (1) development of ecologically clean process equipments for radioactive waste treatment by the HIP method; (2) manufacture of promising protective ceramic compositions and investigation of their physico-mechanical properties

  4. Modeling Hydrogen Generation Rates in the Hanford Waste Treatment and Immobilization Plant

    Energy Technology Data Exchange (ETDEWEB)

    Camaioni, Donald M.; Bryan, Samuel A.; Hallen, Richard T.; Sherwood, David J.; Stock, Leon M.

    2004-03-29

    This presentation describes a project in which Hanford Site and Environmental Management Science Program investigators addressed issues concerning hydrogen generation rates in the Hanford waste treatment and immobilization plant. The hydrogen generation rates of radioactive wastes must be estimated to provide for safe operations. While an existing model satisfactorily predicts rates for quiescent wastes in Hanford underground storage tanks, pretreatment operations will alter the conditions and chemical composition of these wastes. Review of the treatment process flowsheet identified specific issues requiring study to ascertain whether the model would provide conservative values for waste streams in the plant. These include effects of adding hydroxide ion, alpha radiolysis, saturation with air (oxygen) from pulse-jet mixing, treatment with potassium permanganate, organic compounds from degraded ion exchange resins and addition of glass-former chemicals. The effects were systematically investigated through literature review, technical analyses and experimental work.

  5. Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste

    International Nuclear Information System (INIS)

    Koyama, Tadafumi.

    1994-01-01

    A method is described for immobilizing waste chloride salts containing radionuclides such as cesium and strontium and hazardous materials such as barium. A sodalite intermediate is prepared by mixing appropriate amounts of silica, alumina and sodium hydroxide with respect to sodalite and heating the mixture to form the sodalite intermediate and water. Heating is continued to drive off the water to form a water-free intermediate. The water-free intermediate is mixed with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides and hazardous material. The waste salt-intermediate mixture is then compacted and heated under conditions of heat and pressure to form sodalite with the waste salt, radionuclides and hazardous material trapped within the sodalite cage structure. This provides a final product having excellent leach resistant capabilities

  6. Immobilization of metal wastes by reaction with H2S in anoxic basins: concept and elaboration.

    Science.gov (United States)

    Schuiling, R D

    2013-10-01

    Metal wastes are produced in large quantities by a number of industries. Their disposal in isolated waste deposits is certain to cause many subsequent problems, because every material will sooner or later return to the geochemical cycle. The sealing of disposal sites usually starts to leak, often within a short time after the disposal site has been filled. The contained heavy metals are leached from the waste deposit and will contaminate the soil and the groundwater. It is evident that storage as metal sulfides in a permanently anoxic environment is the only safe way to handle metal wastes. The world's largest anoxic basin, the Black Sea, can serve as a georeactor. The metal wastes are sustainably transformed into harmless and immobile solids. These are incorporated in the lifeless bottom muds, where they are stored for millions of years.

  7. Utilization of borosilicate glass for transuranic waste immobilization

    International Nuclear Information System (INIS)

    Ledford, J.A.; Williams, P.M.

    1979-01-01

    Incinerated transuranic waste and other low-level residues have been successfully vitrified by mixing with boric acid and sodium carbonate and heating to 1050 0 C in a bench-scale continuous melter. The resulting borosilicate glass demonstrates excellent mechanical durability and chemical stability

  8. Immobilization of chloride-rich radioactive wastes produced by pyrochemical operations

    International Nuclear Information System (INIS)

    McDaniel, E.W.; Terry, J.W.

    1997-08-01

    A a result of its former role as a producer of nuclear weapons components, the Rocky Flats Environmental Technology Site (RFETS), Golden, Colorado accumulated a variety of plutonium-contaminated materials. When the level of contamination exceeded a predetermined level (the economic discard limit), the materials were classified as residues rather than waste and were stored for later recovery of the plutonium. Although large quantities of residues were processed, others, primarily those more difficult to process, remain in storage at the site. It is planned for the residues with lower concentrations of plutonium to be disposed of as wastes at an appropriate disposal facility, probably the Waste Isolation Pilot Plant (WIPP). Because the plutonium concentration is too high or because the physical or chemical form would be difficult to get into a form acceptable to WIPP, it may not be possible to dispose of a portion of the residues at WIPP. The pyrochemical salts are among the residues that are difficult to dispose of. For a large percentage of the pyrochemical salts, safeguards controls are required, but WIPP was not designed to accommodate safeguards controls. A potential solution would be to immobilize the salts. These immobilized salts would contain substantially higher plutonium concentrations than is currently permissible but would be suitable for disposal at WIPP. This document presents the results of a review of three immobilization technologies to determine if mature technologies exist that would be suitable to immobilize pyrochemical salts: cement-based stabilization, low-temperature vitrification, and polymer encapsulation. The authors recommend that flow sheets and life-cycle costs be developed for cement-based and low-temperature glass immobilization

  9. Lipases Immobilization for Effective Synthesis of Biodiesel Starting from Coffee Waste Oils

    Directory of Open Access Journals (Sweden)

    Lucia Gardossi

    2013-08-01

    Full Text Available Immobilized lipases were applied to the enzymatic conversion of oils from spent coffee ground into biodiesel. Two lipases were selected for the study because of their conformational behavior analysed by Molecular Dynamics (MD simulations taking into account that immobilization conditions affect conformational behavior of the lipases and ultimately, their efficiency upon immobilization. The enzymatic synthesis of biodiesel was initially carried out on a model substrate (triolein in order to select the most promising immobilized biocatalysts. The results indicate that oils can be converted quantitatively within hours. The role of the nature of the immobilization support emerged as a key factor affecting reaction rate, most probably because of partition and mass transfer barriers occurring with hydrophilic solid supports. Finally, oil from spent coffee ground was transformed into biodiesel with yields ranging from 55% to 72%. The synthesis is of particular interest in the perspective of developing sustainable processes for the production of bio-fuels from food wastes and renewable materials. The enzymatic synthesis of biodiesel is carried out under mild conditions, with stoichiometric amounts of substrates (oil and methanol and the removal of free fatty acids is not required.

  10. Study of immobilization of waste from treatment of acid waters of a uranium mining facility

    International Nuclear Information System (INIS)

    Goda, R.T.; Oliveira, A.P. de; Silva, N.C. da; Villegas, R.A.S.; Ferreira, A.M.

    2017-01-01

    This study aimed to produce scientific and technical knowledge aiming at the development of techniques to immobilize the waste generated in the treatment of acid waters in the UTM-INB Caldas uranium mining and processing facility using Portland cement. This residue (calcium diuranate - DUCA) contains uranium compounds and metal hydroxides in a matrix of calcium sulfate. It is observed that this material, in contact with the lake of acid waters of the mine's own pit, undergoes resolubilization and, therefore, changes the quality of the acidic water contained therein, changing the treatment parameters. For the study of immobilization of this residue, the mass of water contained in both the residue deposited in the pit of the mine and in the pulp resulting from the treatment of the acid waters was determined. In addition, different DUCA / CEMENT / WATER ratios were used for immobilization and subsequent mechanical strength and leaching tests. The results showed that in the immobilized samples with 50% cement mass condition, no uranium was detected in the leaching tests, and the mechanical strength at compression was 9.4 MPa, which indicates that more studies are needed, but indicate a good capacity to immobilize uranium in cement

  11. Sampling and analysis plan for the preoperational environmental survey for the immobilized low activity waste (ILAW) project W-465

    International Nuclear Information System (INIS)

    Mitchell, R.M.

    1998-01-01

    This document provides a detailed description of the Sampling and Analysis Plan for the Preoperational Survey to be conducted at the Immobilized Low Activity Waste (ILAW) Project Site in the 200 East Area

  12. Immobilization of defense high-level waste: an assessment of technological strategies and potential regulatory goals. Volume II

    International Nuclear Information System (INIS)

    1979-06-01

    This volume contains the following appendices: selected immobilization processes, directory of selected European organizations involved in HLW management, U.S. high-level waste inventories, and selected European HLW program

  13. Sorption of radioscesium from liquid radioactive waste on clay and immobilization by baking the clay at elevated temperature

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, F.; Ghaffar, A. [Pakistan Inst. of Nuclear Science and Technology, Islamabad (Pakistan)

    2011-07-01

    The cesium-137 is the most problematic radionuclide in the radioactive wastes. It belongs to the IA group of the periodic table, highly reactive towards water and has very high mobility. Due to beta and gamma radiation hazards of radiocesium its decontamination and disposal requires some special tools and techniques. In this study globules of clay material was used for the removal of cesium from low level liquid radioactive wastes and further processed for immobilization. The aim of this study was to assess the solidification and immobilization of secondary waste. The secondary waste, after sorption of cesium from the liquid radioactive waste generated at this institute, was found compatible to the cement matrix used for the cementation process. The procedure for immobilization of low level radioactive waste with cementation using vitreous clay material as an additive was developed. (orig.)

  14. Sorption of radioscesium from liquid radioactive waste on clay and immobilization by baking the clay at elevated temperature

    International Nuclear Information System (INIS)

    Rashid, F.; Ghaffar, A.

    2011-01-01

    The cesium-137 is the most problematic radionuclide in the radioactive wastes. It belongs to the IA group of the periodic table, highly reactive towards water and has very high mobility. Due to beta and gamma radiation hazards of radiocesium its decontamination and disposal requires some special tools and techniques. In this study globules of clay material was used for the removal of cesium from low level liquid radioactive wastes and further processed for immobilization. The aim of this study was to assess the solidification and immobilization of secondary waste. The secondary waste, after sorption of cesium from the liquid radioactive waste generated at this institute, was found compatible to the cement matrix used for the cementation process. The procedure for immobilization of low level radioactive waste with cementation using vitreous clay material as an additive was developed. (orig.)

  15. Modeling a novel glass immobilization waste treatment process using flow

    International Nuclear Information System (INIS)

    Ferrada, J.J.; Nehls, J.W. Jr.; Welch, T.D.; Giardina, J.L.

    1996-01-01

    One option for control and disposal of surplus fissile materials is the Glass Material Oxidation and Dissolution System (GMODS), a process developed at ORNL for directly converting Pu-bearing material into a durable high-quality glass waste form. This paper presents a preliminary assessment of the GMODS process flowsheet using FLOW, a chemical process simulator. The simulation showed that the glass chemistry postulated ion the models has acceptable levels of risks

  16. Compounds with magnetoplumbite or SLNA type structure as materials for nuclear waste immobilization

    International Nuclear Information System (INIS)

    Thery, J.; Vivien, D.; Lejus, A.M.; Collongues, R.

    1985-01-01

    Magnetoplumbite-like structure, and related phases (sodium-lanthanide aluminates: SLNA) are able to accommodate a wide range of elements with various charges and ionic radii. The available coordinences are 4, 5 or 6 for the small cations and 9 or 12 for the large ones. This kind of compounds, which in addition present good chemical inertia, could possibly be used for the immobilization of nuclear waste [fr

  17. River Protection Project (RPP) Immobilized Low- Ativity Waste (ILAW) Disposal Plan

    International Nuclear Information System (INIS)

    BRIGGS, M.G.

    2000-01-01

    This document replaces HNF-1517, Rev 2 which is deleted. It incorporates updates to reflect changes in programmatic direction associated with the vitrification plant contract change and associated DOE/ORP guidance. In addition it incorporates the cancellation of Project W-465, Grout Facility, and the associated modifications to Project W-520, Immobilized High-Level Waste Disposal Facility. It also includes document format changes and section number modifications consistent with CH2M HILL Hanford Group, Inc. procedures

  18. A review of glass-ceramics for the immobilization of nuclear fuel recycle wastes

    International Nuclear Information System (INIS)

    Hayward, P.J.

    1987-01-01

    This report reviews the status of the Canadian, German, U.S., Japanese, U.S.S.R. and Swedish programs for the development of glass-ceramic materials for immobilizing the high-level radioactive wastes arising from the recycling of used nuclear fuel. The progress made in these programs is described, with emphasis on the Canadian program for the development of sphene-based glass-ceramics. The general considerations of product performance and process feasibility for glass-ceramics as a category of waste form material are discussed. 137 refs

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  1. Immobilization of radioactive waste in cement based matrices

    International Nuclear Information System (INIS)

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

    1986-11-01

    Experimental and theoretical studies of hydrated cement systems are described. The behaviour of slag-based cement is described with a view to predicting their long term pH, Esub(n) and mineralogical balance. Modelling studies which enable the prediction at long ages of cement composites are advanced and a base model of the CaO-SiO 2 -H 2 O system presented. The behaviour of U and I in cements is explored. The tolerance of cement systems for a wide range of miscellaneous waste stream components and environmental hazards is described. The redox potential in cements is effectively lowered by irradiation. (author)

  2. Low leach rate glasses for immobilization of nuclear wastes

    International Nuclear Information System (INIS)

    Chick, L.A.; Buckwalter, C.Q.

    1980-10-01

    Improved defense and commercial waste glass have about one order of magnitude lower leach rates at 90 0 C in static deionized water than reference glasses. This durability difference diminishes as the leaching temperature is raised, but at repository temperature less than 150 0 C, the improved compositions would have considerable advantages over reference glases. At the melting temperatures necessary for most of the high-durability glasses, volatility was found to be higher than that experienced in processing current reference glases. Higher volatilities might be compensated for by specific design of the off-gas system for improved off-gas treatment and volatile materials recovery. 6 figures, 2 tables

  3. Sodium zirconium phosphate (NZP) as a host structure for nuclear waste immobilization: A review

    International Nuclear Information System (INIS)

    Scheetz, B.E.; Agrawal, D.K.; Breval, E.; Roy, R.

    1994-01-01

    Sodium zirconium phosphate [NZP] structural family, of which NaZr 2 P 3 O 12 is the parent composition, has been reviewed as a host ceramic waste form for nuclear waste immobilization. NZP compounds are characterized for their ionic conductivity, low thermal expansion and structural flexibility to accommodate a large number of multivalent ions. This latter property of the [NZP] structure allows the incorporation of almost all 42 nuclides present in a typical commercial nuclear waste. The leach studies of simulated waste forms based on NZP have shown reasonable resistance for the release of its constituents. The calculation of dissolution rates of NZP structure has demonstrated that it would take 20,000 times longer to dissolved NZP than quartz

  4. Steam Reforming Technology for Denitration and Immobilization of DOE Tank Wastes

    International Nuclear Information System (INIS)

    Mason, J. B.; McKibbin, J.; Ryan, K.; Schmoker, D.

    2003-01-01

    THOR Treatment Technologies, LLC (THOR) is a joint venture formed in June 2002 by Studsvik, Inc. (Studsvik) and Westinghouse Government Environmental Services Company LLC to further develop, market, and deploy Studsvik's patented THORSM non-incineration, steam reforming waste treatment technology. This paper provides an overview of the THORSM steam reforming process as applied to the denitration and conversion of Department of Energy (DOE) tank wastes to an immobilized mineral form. Using the THORSM steam reforming technology to treat nitrate containing tank wastes could significantly benefit the DOE by reducing capital and life-cycle costs, reducing processing and programmatic risks, and positioning the DOE to meet or exceed its stakeholder commitments for tank closure. Specifically, use of the THORSM technology can facilitate processing of up to 75% of tank wastes without the use of vitrification, yielding substantial life-cycle cost savings

  5. The Remote Handled Immobilization Low Activity Waste Disposal Facility Environmental Permits & Approval Plan

    Energy Technology Data Exchange (ETDEWEB)

    DEFFENBAUGH, M.L.

    2000-08-01

    The purpose of this document is to revise Document HNF-SD-ENV-EE-003, ''Permitting Plan for the Immobilized Low-Activity Waste Project, which was submitted on September 4, 1997. That plan accounted for the interim storage and disposal of Immobilized-Low Activity Waste at the existing Grout Treatment Facility Vaults (Project W-465) and within a newly constructed facility (Project W-520). Project W-520 was to have contained a combination of concrete vaults and trenches. This document supersedes that plan because of two subsequent items: (1) A disposal authorization that was received on October 25, 1999, in a U. S. Department of Energy-Headquarters, memorandum, ''Disposal Authorization Statement for the Department of Energy Hanford site Low-Level Waste Disposal facilities'' and (2) ''Breakthrough Initiative Immobilized Low-Activity Waste (ILAW) Disposal Alternative,'' August 1999, from Lucas Incorporated, Richland, Washington. The direction within the U. S. Department of Energy-Headquarters memorandum was given as follows: ''The DOE Radioactive Waste Management Order requires that a Disposal authorization statement be obtained prior to construction of new low-level waste disposal facility. Field elements with the existing low-level waste disposal facilities shall obtain a disposal authorization statement in accordance with the schedule in the complex-wide Low-Level Waste Management Program Plan. The disposal authorization statement shall be issued based on a review of the facility's performance assessment and composite analysis or appropriate CERCLA documentation. The disposal authorization shall specify the limits and conditions on construction, design, operations, and closure of the low-level waste facility based on these reviews. A disposal authorization statement is a part of the required radioactive waste management basis for a disposal facility. Failure to obtain a disposal authorization statement

  6. Impacts of biochar and oyster shells waste on the immobilization of arsenic in highly contaminated soils.

    Science.gov (United States)

    Chen, Yongshan; Xu, Jinghua; Lv, Zhengyong; Xie, Ruijia; Huang, Liumei; Jiang, Jinping

    2018-07-01

    Soil contamination is a serious problem with deleterious impacts on global sustainability. Readily available, economic, and highly effective technologies are therefore urgently needed for the rehabilitation of contaminated sites. In this study, two readily available materials prepared from bio-wastes, namely biochar and oyster shell waste, were evaluated as soil amendments to immobilize arsenic in a highly As-contaminated soil (up to 15,000 mgAs/kg). Both biochar and oyster shell waste can effectively reduce arsenic leachability in acid soils. After application of the amendments (2-4% addition, w/w), the exchangeable arsenic fraction decreased from 105.8 to 54.0 mg/kg. The application of 2%biochar +2% oyster shell waste most effectively reduced As levels in the column leaching test by reducing the arsenic concentration in the porewater by 62.3% compared with the treatment without amendments. Biochar and oyster shell waste also reduced soluble As(III) from 374.9 ± 18.8 μg/L to 185.9 ± 16.8 μg/L and As(V) from 119.8 ± 13.0 μg/L to 56.4 ± 2.6 μg/L at a pH value of 4-5. The treatment using 4% (w/w) amendments did not result in sufficient As immobilization in highly contaminated soils; high soluble arsenic concentrations (upto193.0 μg/L)were found in the soil leachate, particularly in the form of As(III), indicating a significant potential to pollute shallow groundwater aquifers. This study provides valuable insights into the use of cost-effective and readily available materials for soil remediation and investigates the mechanisms underlying arsenic immobilization in acidic soils. Copyright © 2018 Elsevier Ltd. All rights reserved.

  7. Rapid immobilization of simulated radioactive soil waste by microwave sintering.

    Science.gov (United States)

    Zhang, Shuai; Shu, Xiaoyan; Chen, Shunzhang; Yang, Huimin; Hou, Chenxi; Mao, Xueli; Chi, Fangting; Song, Mianxin; Lu, Xirui

    2017-09-05

    A rapid and efficient method is particularly necessary in the timely disposal of seriously radioactive contaminated soil. In this paper, a series of simulated radioactive soil waste containing different contents of neodymium oxide (3-25wt.%) has been successfully vitrified by microwave sintering at 1300°C for 30min. The microstructures, morphology, element distribution, density and chemical durability of as obtained vitrified forms have been analyzed. The results show that the amorphous structure, homogeneous element distribution, and regular density improvement are well kept, except slight cracks emerge on the magnified surface for the 25wt.% Nd 2 O 3 -containing sample. Moreover, all the vitrified forms exhibit excellent chemical durability, and the leaching rates of Nd are kept as ∼10 -4 -10 -6 g/(m 2 day) within 42days. This demonstrates a potential application of microwave sintering in radioactive contaminated soil disposal. Copyright © 2017 Elsevier B.V. All rights reserved.

  8. Enhanced HLW glass formulations for the waste treatment and immobilization plant

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A. [DOE-WTP Project Office, US Department of Energy, Richland, Washington (United States)

    2013-07-01

    Current estimates and glass formulation efforts are conservative vis-a-vis achievable waste loadings. These formulations have been specified to ensure that glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum, chromium, bismuth, iron, phosphorous, zirconium, and sulfur compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. DOE has a testing program to develop and characterize HLW glasses with higher waste loadings. This work has demonstrated the feasibility of increases in waste loading from 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected these higher waste loading glasses will reduce the HLW canister production requirement by 25% or more. (authors)

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

    International Nuclear Information System (INIS)

    Weingardt, K.M.; Weber, J.R.

    1994-03-01

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

  10. A rapid and continuous system for immobilization of nuclear material waste by vitrification

    International Nuclear Information System (INIS)

    Shareef, M.U.; Hussain, S.H.; Tufail, M.; Rashid, F.

    2009-01-01

    The nuclear technology has prime importance and backbone for rapid development of medical sciences, industries and in power generation as an alternate source of energy. Despite all these facts there is a major problem which is always associated with nuclear technology, that is, the generation of undesirable radioactive wastes. The radioactive wastes are quite problematic and need major attention for its treatment, conditioning and properly disposal to keep the environmental activities and human ecosystem healthy and safe. There are different large scale methods and processes to treat and dispose off the radioactive wastes. These processes are evaluated and designed by the various world competent and pronounced scientists in the light of rules and safety limits set by IAEA and other regulatory authorities to protect the environment and eventually protect our ecosystem. The research and development work on radioactive waste has been proceeding for the last fifty years but still it is a core issue and a big challenge for the nuclear scientists and radiation workers. In this study a rapid and continuous system for immobilization of nuclear waste into glass matrix by vitrification has been designed. In general treatment methods, Borosilicate glass is preferred because it is efficient, cost effective and rapid to that of other radioactive waste form. In this process the simulated waste is mixed with glass forming material and process for melting to form a glassy substrate in continuous manner. The waste is being converted into vitreous form and encapsulate into a glass matrix. (author)

  11. Technetium Incorporation in Glass for the Hanford Tank Waste Treatment and Immobilization Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Kim, Dong Sang

    2015-01-14

    A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of 99Tc (~9 × 10E2 TBq or ~2.5 × 104 Ci or ~1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with 99Tc is its high mobility in addition to a long half-life (2.1×105 yrs). The pertechnetate ion (TcO4-) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater

  12. Immobilized low-activity waste site borehole 299-E17-21

    International Nuclear Information System (INIS)

    Reidel, S.P.; Reynolds, K.D.; Horton, D.G.

    1998-08-01

    The Tank Waste Remediation System (TWRS) is the group at the Hanford Site responsible for the safe underground storage of liquid waste from previous Hanford Site operations, the storage and disposal of immobilized tank waste, and closure of underground tanks. The current plan is to dispose of immobilized low-activity tank waste (ILAW) in new facilities in the southcentral part of 200-East Area and in four existing vaults along the east side of 200-East Area. Boreholes 299-E17-21, B8501, and B8502 were drilled at the southwest corner of the ILAW site in support of the Performance Assessment activities for the disposal options. This report summarizes the initial geologic findings, field tests conducted on those boreholes, and ongoing studies. One deep (480 feet) borehole and two shallow (50 feet) boreholes were drilled at the southwest corner of the ILAW site. The primary factor dictating the location of the boreholes was their characterization function with respect to developing the geohydrologic model for the site and satisfying associated Data Quality Objectives. The deep borehole was drilled to characterize subsurface conditions beneath the ILAW site, and two shallow boreholes were drilled to support an ongoing environmental tracer study. The tracer study will supply information to the Performance Assessment. All the boreholes provide data on the vadose zone and saturated zone in a previously uncharacterized area

  13. The used epoxy matrix in immobilization sludge process of alpha emitter radioactive waste

    International Nuclear Information System (INIS)

    Walman, E.; Salimin, Z.; Johan, B.

    1998-01-01

    Immobilization of alpha emitter radioactive waste containing of ion complex of uranyl carbonate on uranium concentration ≤ 50 mg/l has been carried out using epoxy matrix. The first step of process is the coagulation of uranium with 1.3 mole/l of Ca(OH) 2 coagulant concentration on pH 8 to precipitate the calcium uranyl carbonate on uranium concentration ≤ g/l. The immobilization of calcium uranyl carbonate with epoxy matrix was done on variation of the ratio of resin epoxy and hardener of 1 : 1 (giving the maximum value of density and compressive strength), the increasing of precipitate loading capacity give the decreasing of compressive strength of embedded waste. The test of compressive strength and leaching was done for the embedded waste after its curing time using Paul Weber equipment and 7 days immersion of samples in normal water. On the precipitate loading capacity of 70%, the quality of embedded waste still conform to the standard quality value i.e. density 1.2 g/cm 3 , compressive strength 10 kN/cm 2 and there is not any release of radionuclide during leaching test (undetectable).. (author)

  14. Glass as a waste form for the immobilization of plutonium

    International Nuclear Information System (INIS)

    Bates, J.K.; Ellison, A.J.G.; Emery, J.W.; Hoh, J.C.

    1995-01-01

    Several alternatives for disposal of surplus plutonium are being considered. One method is incorporating Pu into glass and in this paper we discuss the development and corrosion behavior of an alkali-tin-silicate glass and update results in testing Pu doped Defense Waste Processing Facility (DWPF) reference glasses. The alkali-tin-silicate glass was engineered to accommodate a high Pu loading and to be durable under conditions likely to accelerate glass reaction. The glass dissolves about 7 wt% Pu together with the neutron absorber Gd, and under test conditions expected to accelerate the glass reaction with water, is resistant to corrosion. The Pu and the Gd are released from the glass at nearly the same rate in static corrosion tests in water, and are not segregated into surface alteration phases when the glass is reacted in water vapor. Similar results for the behavior of Pu and Gd are found for the DWPF reference glasses, although the long-term rate of reaction for the reference glasses is more rapid than for the alkali-tin-silicate glass

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

    International Nuclear Information System (INIS)

    Shi, C.; Day, R.L.

    1996-01-01

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

  16. Magnesium Potassium Phosphate Compound for Immobilization of Radioactive Waste Containing Actinide and Rare Earth Elements

    Directory of Open Access Journals (Sweden)

    Sergey E. Vinokurov

    2018-06-01

    Full Text Available The problem of effective immobilization of liquid radioactive waste (LRW is key to the successful development of nuclear energy. The possibility of using the magnesium potassium phosphate (MKP compound for LRW immobilization on the example of nitric acid solutions containing actinides and rare earth elements (REE, including high level waste (HLW surrogate solution, is considered in the research work. Under the study of phase composition and structure of the MKP compounds that is obtained by the XRD and SEM methods, it was established that the compounds are composed of crystalline phases—analogues of natural phosphate minerals (struvite, metaankoleite. The hydrolytic stability of the compounds was determined according to the semi-dynamic test GOST R 52126-2003. Low leaching rates of radionuclides from the compound are established, including a differential leaching rate of 239Pu and 241Am—3.5 × 10−7 and 5.3 × 10−7 g/(cm2∙day. As a result of the research work, it was concluded that the MKP compound is promising for LRW immobilization and can become an alternative material combining the advantages of easy implementation of the technology, like cementation and the high physical and chemical stability corresponding to a glass-like compound.

  17. Foaming and Antifoaming and Gas Entrainment in Radioactive Waste Pretreatment and Immobilization Processes. Final Report

    International Nuclear Information System (INIS)

    Wasan, Darsh T.

    2007-01-01

    The Savannah River Site (SRS) and Hanford site are in the process of stabilizing millions of gallons of radioactive waste slurries remaining from production of nuclear materials for the Department of Energy (DOE). The Defense Waste Processing Facility (DWPF) at SRS is currently vitrifying the waste in borosilicate glass, while the facilities at the Hanford site are in the construction phase. Both processes utilize slurry-fed joule-heated melters to vitrify the waste slurries. The DWPF has experienced difficulty during operations. The cause of the operational problems has been attributed to foaming, gas entrainment and the rheological properties of the process slurries. The rheological properties of the waste slurries limit the total solids content that can be processed by the remote equipment during the pretreatment and meter feed processes. Highly viscous material can lead to air entrainment during agitation and difficulties with pump operations. Excessive foaming in waste evaporators can cause carryover of radionuclides and non-radioactive waste to the condensate system. Experimental and theoretical investigations of the surface phenomena, suspension rheology and bubble generation of interactions that lead to foaming and air entrainment problems in the DOE High Level and Low Activity Radioactive Waste separation and immobilization processes were pursued under this project. The first major task accomplished in the grant proposal involved development of a theoretical model of the phenomenon of foaming in a three-phase gas-liquid-solid slurry system. This work was presented in a recently completed Ph.D. thesis (9). The second major task involved the investigation of the inter-particle interaction and microstructure formation in a model slurry by the batch sedimentation method. Both experiments and modeling studies were carried out. The results were presented in a recently completed Ph.D. thesis. The third task involved the use of laser confocal microscopy to study

  18. Borosilicate glass as a matrix for immobilization of SRP high-level waste

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1980-01-01

    Approximately 22 million gallons of high-level radioactive defense waste are currently being stored in large underground tanks located on the Savannah River Plant (SRP) site in Aiken, South Carolina. One option now being considered for long-term management of this waste involves removing the waste from the tanks, chemically processing the waste, and immobilizing the potentially harmful radionuclides in the waste into a borosilicate glass matrix. The technology for producing waste glass forms is well developed and has been demonstrated on various scales using simulated as well as radioactive SRP waste. Recently, full-scale prototypical equipment has been made operational at SRP. This includes both a joule-heated ceramic melter and an in-can melter. These melters are a part of an integrated vitrification system which is under evaluation and includes a spray calciner, direct liquid feed apparatus, and various elements of an off-gas system. Two of the most important properties of the waste glass are mechanical integrity and leachability. Programs are in progress at SRL aimed at minimizing thermally induced cracking by carefully controlling cooling cycles and using ceramic liners or coatings. The leachability of SRP waste glass has been studied under many different conditions and consistently found to be low. For example, the leachability of actual SRP waste glass was found to be 10 -6 to 10 -5 g/(cm 2 )(day) initially and decreasing to 10 -9 to 10 -8 g/(cm 2 )(day) after 100 days. Waste glass is also being studied under anticipated storage conditions. In brine at 90 0 C, the leachability is about 5 x 10 -8 g/(cm 2 )(day) after 60 days. The effects of other geological media including granite, basalt, shale, and tuff are also being studied as part of the multibarrier isolation system

  19. Development of crystalline ceramic for immobilization of TRU wastes in V.G. Khlopin Radium Institute

    International Nuclear Information System (INIS)

    Burakov, B.E.; Anderson, E.B.

    1999-01-01

    This paper discusses the Radium Institute's experience in the synthesis of crystalline ceramics based on two groups of actinide host-phases: 1) Zircon/zirconia-(Zn, Ac)SiO 4 /(Zr, Ac)O 2 , where Ac=Pu, Np, Am, Cm; 2) Garnet/perovskite-(Y, Gd, Ac) 3 (Al, Ga, Ac,..) 5 O 12 /(Y, Gd, Ac)(Al, Ga)O 3 . The zircon/zirconia ceramic was suggested as an universal waste form for the immobilization of TRU as well as weapon-grade Pu. Because the position of the Russian Ministry of Atomic Energy (Minatom) does not consider weapons Pu as a waste', the Radium Institute proposed the use of the same ceramic (mainly monophase zirconia ) as a Pu-fuel. The garnet/perovskite ceramic was suggested for the immobilization of military TRU wastes of complex chemical composition. The advantage of this ceramic is that Garnet and Perovskite host-phases can incorporate in their lattices not only actinides, but also other elements including neutron absorbers in a broad range of concentration and in different valence state. Sample of zircon/zirconia ceramic were prepared by hot uniaxial pressing (at temperature T=1300, 1400, 1500degC and pressure P=25 MPa) and sintering (at T=1450, 1490, 1500, 1600degC) methods using different types of initial precursor. Samples of garnet/perovskite ceramic were synthesized by melting method at T=2000degC. Ce, U, Gd were used as TRU stimulants for both types of ceramic. One sample of zircon/zirconia ceramic was doped with 10 wt.% of Pu 239 . Physico-chemical features of these ceramics are described. In conclusion we propose that the pressureless technology based on sintering or melting methods be used for the synthesis of ceramics for the immobilization of all types of TRU wastes. (author)

  20. Immobilization of Technetium Waste from Pyro-processing Using Tellurite Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Jong; Pyo, Jae-Young; Lee, Cheong-Won [POSTECH, Pohang (Korea, Republic of); Yang, Jae-Hwan; Park, Hwan-Seo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Vitrification of Tc wastes has been challenging because of the low solubility in the silicate glass and high volatility in the melting process. In previous studies, the measured solubility of Tc and Re was ⁓ 3000 ppm at 1000 .deg. C in low activity waste (LAW) glass. And retention of Tc has been reported within 12 - 77% during the borosilicate vitrification process. Tellurite glasses have been studied for halide waste immobilization due to low melting temperatures (Tm= 600-800 .deg. C) and flexibility of network with foreign ions. Tellurite glasses offered higher halide retention than borosilicate glasses. The structure of pure tellurite (TeO{sub 2}) consists of TeO{sub 4} trigonal bipyramids (tbp), but TeO{sub 4} units are converted to TeO{sub 3} trigonal pyramids (tp) having non-bridging oxygen (NBO) as the modifiers added. Objectives of this study are to investigate the tellurite glasses for Tc immobilization using Re as a surrogate. Retention and waste loading of Re were analyzed during the vitrification process of tellurite glass. We investigated local structures of Re ions in glasses by Raman and X-ray absorption spectroscopies. The tellurite glass was investigated to immobilize the Ca(TcO{sub 4}){sub 2}, surrogated by Ca(ReO{sub 4}){sub 2}. The average of Re retention in tellurite glass was 86%. The 7-day PCT results were satisfied with U.S requirement up to 9 mass% of Ca(ReO{sub 4}){sub 2} content. Re in the tellurite glass exists +7 oxidation state and was coordinated with 4 oxygen.

  1. A Strategy for Maintenance of the Long-Term Performance Assessment of Immobilized Low-Activity Waste Glass

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, Joseph V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Freedman, Vicky L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-28

    Approximately 50 million gallons of high-level radioactive mixed waste has accumulated in 177 buried single- and double-shell tanks at the Hanford Site in southeastern Washington State as a result of the past production of nuclear materials, primarily for defense uses. The United States Department of Energy (DOE) is proceeding with plans to permanently dispose of this waste. Plans call for separating the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, which will be vitrified at the Hanford Waste Treatment and Immobilization Plant (WTP). Principal radionuclides of concern in LAW are 99Tc, 129I, and U, while non-radioactive contaminants of concern are Cr and nitrate/nitrite. HLW glass will be sent off-site to an undetermined federal site for deep geological disposal while the much larger volume of immobilized low-activity waste will be placed in the on-site, near-surface Integrated Disposal Facility (IDF).

  2. Characterization and leach investigations of sodium silicate matrices used for immobilization of radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Sharaf El-Deen, A N; El-Dessouky, M M; Helmy, M A [Petroleum Research Institue, Academy of Scientific Research, Nasr City, Cairo (Egypt); Abed Raouf, M W; El-Dessouky, M I [Hot Lab. Centre, Atomic Energy Authority, Cairo (Egypt)

    1995-10-01

    In this study, simulated liquid waste and radioactive tracers of Cs-137 and Co-60 were used to represent the high-level liquid waste (HLLW). immobilization of the liquid waste was performed by its interaction with commercial sodium silicate hydrosol to the gel point, at room temperature. The candidate waste forms forms were fabricated from the obtained hydrogel through several steps including: drying the hydrogel to a solid gel form, crushing the solid to be in a powder from, pressing the powder to the green disk form using a cold pressing technique and finally the heat treatment of the green disks to the sintered form. Characterization for the obtained waste forms was carried out using: thermal analysis (TGA and DTA), X-ray powder diffraction (XRD) techniques and porosity investigation. The leach tests for the prepared forms were conducted according to the international atomic energy agency (IAEA) standard test (static and accelerated). The static test was carried out for simulated and radioactive waste in distilled, bidistilled and ground water for 28 days. The accelerated (Soxhlet) test was conducted for simulated waste in deionized water for 72 hours. 4 figs., 7 tabs.

  3. Waste incineration and immobilization for nuclear facilities. Status report, April-September 1978

    International Nuclear Information System (INIS)

    Johnson, A.J.; Williams, P.M.; Burkhardt, S.C.; Ledford, J.A.; Gallagher, K.Y.

    1980-01-01

    The fluidized bed incinerator and waste immobilization processes are being developed to process various liquid and solid wastes that are generated by a nuclear facility. The versatility of the incinerator liquid waste handling system has been enhanced by recent changes made in the pumping and related piping system. Tributyl phosphate-solvent incineration has been evaluated thoroughly using the pilot plant fluidized bed incinerator. Vitrified glass pellets were made to determine operating parameters of a resistance-heated reactor and to produce samples for testing. Procedures were developed for testing the product pellets. A simplified start-up procedure was devised as development continued on a second type of reactor, the Joule-heated melter

  4. Enzymatic saccharification of Tapioca processing wastes into biosugars through immobilization technology (Mini Review

    Directory of Open Access Journals (Sweden)

    Nurul Aini Edama

    2014-03-01

    Full Text Available Cassava is very popular in Nigeria, Brazil, Thailand and Indonesia. The global cassava production is currently estimated at more than 200 million tons and the trend is increasing due to higher demand for food products. Together with food products, huge amounts of cassava wastes are also produced including cassava pulp, peel and starchy wastewater. To ensure the sustainability of this industry, these wastes must be properly managed to reduce serious threat to the environment and among the strategies to achieve that is to convert them into biosugars. Later on, biosugars could be converted into other end products such as bioethanol. The objective of this paper is to highlight the technical feasibility and potentials of converting cassava processing wastes into biosugars by understanding their generation and mass balance at the processing stage. Moreover, enzyme immobilization technology for better biosugar conversion and future trends are also discussed.

  5. Design requirements document for Project W-465, immobilized low-activity waste interim storage

    International Nuclear Information System (INIS)

    Burbank, D.A.

    1998-01-01

    The scope of this Design Requirements Document (DRD) is to identify the functions and associated requirements that must be performed to accept, transport, handle, and store immobilized low-activity waste (ILAW) produced by the privatized Tank Waste Remediation System (TWRS) treatment contractors. The functional and performance requirements in this document provide the basis for the conceptual design of the TWRS ILAW Interim Storage facility project and provides traceability from the program level requirements to the project design activity. Technical and programmatic risk associated with the TWRS planning basis are discussed in the Tank Waste Remediation System Decisions and Risk Assessment (Johnson 1994). The design requirements provided in this document will be augmented by additional detailed design data documented by the project

  6. Supplemental Immobilization of Hanford Low-Activity Waste: Cast Stone Augmented Formulation Matrix Tests

    International Nuclear Information System (INIS)

    Cozzi, A.; Crawford, C.; Fox, K.; Hansen, E.; Roberts, K.

    2015-01-01

    More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy's (DOE's) Hanford Site in Washington State. The HLW will be vitrified in the HLW facility for ultimate disposal at an offsite federal repository. A portion (~35%) of the LAW will be vitrified in the LAW vitrification facility for disposal onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize all of the wastes destined for those facilities. However, a second facility will be needed for the expected volume of LAW requiring immobilization. Cast Stone, a cementitious waste form, is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with the waste acceptance criteria for the disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. A testing program was developed in fiscal year (FY) 2012 describing in detail the work needed to develop and qualify Cast Stone as a waste form for the solidification of Hanford LAW. A statistically designed test matrix was used to evaluate the effects of key parameters on the properties of the Cast Stone as it is initially prepared and after curing. For the processing properties, the water-to-dry-blend mix ratio was the most significant parameter in affecting the range of values observed for each property. The single shell tank (SST) Blend simulant also showed differences in measured properties compared to the other three simulants tested. A review of the testing matrix and results indicated that an additional set of tests would be beneficial to improve the understanding of the impacts noted in the

  7. Supplemental Immobilization of Hanford Low-Activity Waste: Cast Stone Augmented Formulation Matrix Tests

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fox, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hansen, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Roberts, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-20

    More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in Washington State. The HLW will be vitrified in the HLW facility for ultimate disposal at an offsite federal repository. A portion (~35%) of the LAW will be vitrified in the LAW vitrification facility for disposal onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize all of the wastes destined for those facilities. However, a second facility will be needed for the expected volume of LAW requiring immobilization. Cast Stone, a cementitious waste form, is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with the waste acceptance criteria for the disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. A testing program was developed in fiscal year (FY) 2012 describing in detail the work needed to develop and qualify Cast Stone as a waste form for the solidification of Hanford LAW. A statistically designed test matrix was used to evaluate the effects of key parameters on the properties of the Cast Stone as it is initially prepared and after curing. For the processing properties, the water-to-dry-blend mix ratio was the most significant parameter in affecting the range of values observed for each property. The single shell tank (SST) Blend simulant also showed differences in measured properties compared to the other three simulants tested. A review of the testing matrix and results indicated that an additional set of tests would be beneficial to improve the understanding of the impacts noted in the Screening

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

    International Nuclear Information System (INIS)

    Sami, N.M.

    2013-01-01

    The problem of radioactive waste management has been raised from the beginning use of nuclear energy for different purposes. The rad waste streams produced were sufficient to cause dangerous effects to man and its environment. The ordinary portland cement is the material more extensively used in the technologies of solidification and immobilization of the toxic wastes, low and medium level radioactive wastes. The production of portland cement is one of the most energy-intensive and polluting. The use of high energy in the production causes high emission due to the nature and processes of raw materials. The cement industry is responsible for 7% of the total CO 2 emission. Thus, the cement industry has a crucial role in the global warming. The formation of alite (Ca 3 SiO 5 ), which is the main component of the Portland cement clinker, produces a greater amount of CO 2 emission than the formation of belite (Ca 2 SiO 4 ). The proportion of alite to belite is about 3 in ordinary Portland clinker. Therefore, by decreasing this proportion less CO 2 would be emitted. Furthermore, if industrial byproducts such as fly ash from thermal power station or from incineration of municipal solid wastes have the potential to reduce CO 2 used as raw materials and alternative hydrothermal calcination routes are employed for belite clinker production, CO 2 emission can be strongly reduced or even totally avoided. The availability of fly ash will help in reducing the CO 2 emissions and will also help in resolving, to a great extent, the fly ash disposal problem. This thesis is based on focusing on the possibility of using fly ash as raw materials to prepare low cost innovation matrices for immobilization of radioactive wastes by synthesizing new kind of cement of low consuming energy. The synthesis process is based on the hydrothermal-calcination route of the fly ash without extra additions.

  9. Separation, Concentration, and Immobilization of Technetium and Iodine from Alkaline Supernate Waste

    Energy Technology Data Exchange (ETDEWEB)

    James Harvey; Michael Gula

    1998-12-07

    Development of remediation technologies for the characterization, retrieval, treatment, concentration, and final disposal of radioactive and chemical tank waste stored within the Department of Energy (DOE) complex represents an enormous scientific and technological challenge. A combined total of over 90 million gallons of high-level waste (HLW) and low-level waste (LLW) are stored in 335 underground storage tanks at four different DOE sites. Roughly 98% of this waste is highly alkaline in nature and contains high concentrations of nitrate and nitrite salts along with lesser concentrations of other salts. The primary waste forms are sludge, saltcake, and liquid supernatant with the bulk of the radioactivity contained in the sludge, making it the largest source of HLW. The saltcake (liquid waste with most of the water removed) and liquid supernatant consist mainly of sodium nitrate and sodium hydroxide salts. The main radioactive constituent in the alkaline supernatant is cesium-137, but strontium-90, technetium-99, and transuranic nuclides are also present in varying concentrations. Reduction of the radioactivity below Nuclear Regulatory Commission (NRC) limits would allow the bulk of the waste to be disposed of as LLW. Because of the long half-life of technetium-99 (2.1 x 10 5 y) and the mobility of the pertechnetate ion (TcO 4 - ) in the environment, it is expected that technetium will have to be removed from the Hanford wastes prior to disposal as LLW. Also, for some of the wastes, some level of technetium removal will be required to meet LLW criteria for radioactive content. Therefore, DOE has identified a need to develop technologies for the separation and concentration of technetium-99 from LLW streams. Eichrom has responded to this DOE-identified need by demonstrating a complete flowsheet for the separation, concentration, and immobilization of technetium (and iodine) from alkaline supernatant waste.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

    Low and Intermediate level radioactive wastes (LILW) and hazardous wastes, presents a waste disposal problem. In this respect, a process to immobilize different radioactive and hazardous wastes, including metals contaminated with radionuclides in a form that is non-dispersible and meet the Environmental Protection Agency (USA, EPA) leaching criteria is a must. In this stabilization and solidification process (S/S), simulated radioactive wastes of Cs, Sr, Ce, Cr, and Pb were immobilized in the stable form of sulfur polymer cement (SPC). In the present work, the mixture of the contaminant(s) and the sulfur mixture which is composed from 95% S and 5% aromatic/or aliphatic hydrocarbons used as polymerizing agents for sulfur (by weight), were added in a stainless steel vessel and primarily heated to 40 degree C for four hours, this time was sufficient for homogeneous mixing of the metals with sulfur and Na 2 S (to convert the metals to their corresponding sulfides). Additional SPC was then added and the temperature of the mixture was raised to 135 ±1 degree C, resulting in a molten form that was poured into a stainless steel mold where it cooled and solidified. Durability of the fabricated SPC matrices was assessed in terms of water of immersion, porosity, and compressive strength. The water absorption and open porosity were very low and didn't exceed 2.5 % for all matrices, whereas the compressive strength ranged between 7 and 14 KN/m 2 depending on the matrix composition. The immobilized waste forms of SPC were characterized by X-ray diffraction (XRD) technique that showed that the different contaminants were stabilized during the solidification process to form stable sulfides. Leachability of the waste matrices was assessed by the Toxicity Characteristic Leaching Procedure (TCLP) of the EPA, optimized and compared with the new EPA Universal Treatment Standards.The TCLP results showed that the concentration of the most contaminants released were under detection limit

  11. Managing the process for storage and disposal of immobilized high- and low-level tank waste at the Hanford Site

    International Nuclear Information System (INIS)

    Murkowski, R.J.

    1998-01-01

    Lockheed Martin Hanford Corporation (LMHC) is one of six subcontractors under Fluor Daniel Hanford, Inc., the Management and Integration contractor for the Project Hanford Management Contract working for the US Department of Energy. One of LMHC's responsibilities is to prepare storage and disposal facilities to receive immobilized high and low-level tank waste by June of 2002. The immobilized materials are to be produced by one or more vendors working under a privatization contract. The immobilized low-activity waste is to be permanently disposed of at the Hanford Site while the immobilized high-level waste is to be stored at the Hanford Site while awaiting shipment to the offsite repository. Figure 1 is an overview of the entire cleanup mission with the disposal portion of the mission. Figure 2 is a representation of major activities required to complete the storage and disposal mission. The challenge for the LNIHC team is to understand and plan for accepting materials that are described in the Request for Proposal. Private companies will submit bids based on the Request for Proposal and other Department of Energy requirements. LMHC, however, must maintain sufficient flexibility to accept modifications that may occur during the privatization bid/award process that is expected to be completed by May 1998. Fundamental to this planning is to minimize the risks of stand-by costs if storage and disposal facilities are not available to receive the immobilized waste. LMHC has followed a rigorous process for the identification of the functions and requirements of the storage/disposal facilities. A set of alternatives to meet these functions and requirements were identified and evaluated. The alternatives selected were (1) to modify four vaults for disposal of immobilized low-activity waste, and (2) to retrofit a portion of the Canister Storage Building for storage of immobilized high-level waste

  12. DEVELOPMENT, QUALIFICATION, AND DISPOSAL OF AN ALTERNATIVE IMMOBILIZED LOW-ACTIVITY WASTE FORM AT THE HANFORD SITE

    International Nuclear Information System (INIS)

    Sams, T.L.; Edge, J.A.; Swanberg, D.J.; Robbins, R.A.

    2011-01-01

    Demonstrating that a waste form produced by a given immobilization process is chemically and physically durable as well as compliant with disposal facility acceptance criteria is critical to the success of a waste treatment program, and must be pursued in conjunction with the maturation of the waste processing technology. Testing of waste forms produced using differing scales of processing units and classes of feeds (simulants versus actual waste) is the crux of the waste form qualification process. Testing is typically focused on leachability of constituents of concern (COCs), as well as chemical and physical durability of the waste form. A principal challenge regarding testing immobilized low-activity waste (ILAW) forms is the absence of a standard test suite or set of mandatory parameters against which waste forms may be tested, compared, and qualified for acceptance in existing and proposed nuclear waste disposal sites at Hanford and across the Department of Energy (DOE) complex. A coherent and widely applicable compliance strategy to support characterization and disposal of new waste forms is essential to enhance and accelerate the remediation of DOE tank waste. This paper provides a background summary of important entities, regulations, and considerations for nuclear waste form qualification and disposal. Against this backdrop, this paper describes a strategy for meeting and demonstrating compliance with disposal requirements emphasizing the River Protection Project (RPP) Integrated Disposal Facility (IDF) at the Hanford Site and the fluidized bed steam reforming (FBSR) mineralized low-activity waste (LAW) product stream.

  13. Study of immobilization of radioactive wastes in asphaltic matrices and elastomeric residues by using microwave technique

    International Nuclear Information System (INIS)

    Caratin, Reinaldo Leonel

    2007-01-01

    In the present work, the technique of microwave heating was used to study the immobilization of low and intermediate activity level radioactive waste, such as spent ion exchange resin used to remove undesirable ions of primary circuits of refrigeration in water refrigerated nuclear reactors, and those used in chemical and radionuclide separation columns in the quality control of radioisotopes. Bitumen matrices reinforced with some kinds of rubber (Neoprene R , silicon and ethylene-vinyl-acetate), from production leftovers or scraps, were used for incorporation of radioactive waste. The samples irradiation was made in a home microwave oven that operates at a frequency of 2.450 MHZ with 1.000 W power. The samples were characterized by developing assays on penetration, leaching resistance, softening, flash and combustion points, thermogravimetry and optical microscopy. The obtained results were compatible with the pattern of matrices components, which shows that technique is a very useful alternative to conventional immobilization methods and to those kinds of radioactive waste. (author)

  14. Treatment of waste salt from the advanced spent fuel conditioning process (II) : optimum immobilization condition

    International Nuclear Information System (INIS)

    Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

    2004-01-01

    Since zeolite is known to be stable at a high temperature, it has been reported as a promising immobilization matrix for waste salt. The crystal structure of dehydrated zeolite A breaks down above 1060 K, resulting in the formation of an amorphous solid and re-crystallization to beta-Cristobalite. This structural degradation depends on the existence of chlorides. When contacted to HCl, zeolite 4A is not stable even at 473 K. The optimum consolidation condition for LiCl salt waste from the oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) has been studied using zeolite A since 2001. Actually the constituents of waste salt are water-soluble. And, alkali halides are known to be readily radiolyzed to yield interstitial halogens and metal colloids. For disposal in a geological repository, the waste salt must meet the acceptance criteria. For a waste form containing chloride salt, two of the more important criteria are leach resistance and waste form durability. In this work, we prepared some samples with different mixing ratios of LiCl salt to zeolite A, and then compared some characteristics such as thermal stability, salt occlusion, free chloride content, leach resistance, mixing effect, etc

  15. Immobilization of low and intermediate level radioactive liquid wastes using some industrial by-product materials

    International Nuclear Information System (INIS)

    Sami, N.M.; EI-Dessouky, M.I.; Abou EI-Nour, F.H.; Abdel-Khalik, M.

    2006-01-01

    Immobilization of low and intermediate level.radioactive liquid wastes in different matrices: ordinary Portland cement and cement mixed with some industrial byproduct: by-pass kiln cement dust, blast furnace slag and ceramic sludge was studied. The effect of these industrial by-product materials on the compressive strength, water immersion, radiation effect and teachability were investigated. The obtained results showed that, these industrial by-product improve the cement pastes where they increase the compressive strength, decrease the leaching rate for radioactive cesium-137 and cobalt-60 ions through the solidified waste forms and increase resistance for y-radiation. It is found that, solidified waste forms of intermediate level liquid waste (ILLW) had high compressive strength values more than those obtained from low level liquid waste (LLLW). The compressive strength increased after immersion in different leachant for one and three months for samples with LLLW higher than those obtained for ILLW. The cumulative fractions released of cesium-137 and cobalt-60 of solidified waste forms of LLLW was lower than those obtained for ILLW

  16. Preliminary evaluation of alternative forms for immobilization of Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Stone, J.A.; Goforth, S.T. Jr.; Smith, P.K.

    1979-12-01

    An evaluation of available information on eleven alternative solid forms for immobilization of SRP high-level waste has been completed. Based on the assessment of both product and process characteristics, four forms were selected for more detailed evaluation: (1) borosilicate glass made in the reference process, (2) a high-silica glass made from a porous glass matrix, (3) crystalline ceramics such as supercalcine or SYNROC, and (4) ceramics coated with an impervious barrier. The assessment includes a discussion of product and process characteristics for each of the eleven forms, a cross comparison of these characteristics for the forms, and the bases for selecting the most promising forms for further study

  17. Sintered bentonite ceramics for the immobilization of cesium- and strontium-bearing radioactive waste

    Science.gov (United States)

    Ortega, Luis Humberto

    The Advanced Fuel Cycle Initiative (AFCI) is a Department of Energy (DOE) program, that has been investigating technologies to improve fuel cycle sustainability and proliferation resistance. One of the program's goals is to reduce the amount of radioactive waste requiring repository disposal. Cesium and strontium are two primary heat sources during the first 300 years of spent nuclear fuel's decay, specifically isotopes Cs-137 and Sr-90. Removal of these isotopes from spent nuclear fuel will reduce the activity of the bulk spent fuel, reducing the heat given off by the waste. Once the cesium and strontium are separated from the bulk of the spent nuclear fuel, the isotopes must be immobilized. This study is focused on a method to immobilize a cesium- and strontium-bearing radioactive liquid waste stream. While there are various schemes to remove these isotopes from spent fuel, this study has focused on a nitric acid based liquid waste. The waste liquid was mixed with the bentonite, dried then sintered. To be effective sintering temperatures from 1100 to 1200°C were required, and waste concentrations must be at least 25 wt%. The product is a leach resistant ceramic solid with the waste elements embedded within alumino-silicates and a silicon rich phase. The cesium is primarily incorporated into pollucite and the strontium into a monoclinic feldspar. The simulated waste was prepared from nitrate salts of stable ions. These ions were limited to cesium, strontium, barium and rubidium. Barium and rubidium will be co-extracted during separation due to similar chemical properties to cesium and strontium. The waste liquid was added to the bentonite clay incrementally with drying steps between each addition. The dry powder was pressed and then sintered at various temperatures. The maximum loading tested is 32 wt. percent waste, which refers to 13.9 wt. percent cesium, 12.2 wt. percent barium, 4.1 wt. percent strontium, and 2.0 wt. percent rubidium. Lower loadings of waste

  18. Immobilization of aqueous radioactive cesium wastes by conversion to aluminosilicate minerals

    International Nuclear Information System (INIS)

    Barney, G.S.

    1975-05-01

    Radioactive cesium (primarily 137 Cs) is a major toxic constituent of liquid wastes from nuclear fuel processing plants. Because of the long half-life, highly penetrating radiation, and mobility of 137 Cs, it is necessary to convert wastes containing this radioisotope into a solid form which will prevent movement to the biosphere during long-term storage. A method for converting cesium wastes to solid, highly insoluble, thermally stable aluminosilicate minerals is described. Aluminum silicate clays (bentonite, kaolin, or pyrophyllite) or hydrous aluminosilicate gels are reacted with basic waste solutions to form pollucite, cesium zeolite (Cs-D), Cs-F, cancrinite, or nepheline. Cesium is trapped in the aluminosilicate crystal lattice of the mineral and is permanently immobilized. The identity of the mineral product is dependent on the waste composition and the SiO 2 /Al 2 O 3 ratio of the clay or gel. The stoichiometry and kinetics of mineral formation reactions are described. The products are evaluated with respect to leachability, thermal stability, and crystal morphology. (U.S.)

  19. Studies on gelation of sodium silicate hydrosol for immobilization of high level liquid waste (HLLW).

    Energy Technology Data Exchange (ETDEWEB)

    Abdel Raouf, M W [Hot Lab. Centre, Atomic Energy Authority, Cairo (Egypt); Sharaf El-deen, A N; El-Dessouky, M M [Military Technical College, Kobry El-Kobbah, Cairo (Egypt)

    1995-10-01

    Immobilization of the simulated high-level liquid waste (HLLW) was performed via the gelation with sodium silicate hydrosol at room temperature. The simulated waste in this study, was represented by the electrolytes of Li, Na, K, Cs, Co and Sr at different concentrations. Specific loading of the liquid waste with 0.6 M Mg (NO{sub 3})2 and tailoring with Al salts were tried during most of the gelation processes. Mineral acid (HCl or {sub 3}) were added during the gelation processes to achieve the gel point, especially when lower concentrations of the simulated waste were used. The obtained hydrogel were dried to obtain the solid gel form. The gelation processes were investigated in terms of the different factors that affected them, namely: temperature, pH, changes in the concentration of the initial hydrosol and the used electrolytes. The efficiency of the gelation processes was investigated from the ratio of the amount of simulated waste reacted (m mole) to the initial silicate used (m mole), i.e. X value. Lower X values were observed when using multi valent cations (higher polarizing power). A special effect of increasing the sorption of metal cations in the silica matrix was observed when Al{sup 3+} replaced Si{sup 4+} in the three-dimensional network structure of the matrix. 3 figs., 7 tabs.

  20. Liquidus temperature and chemical durability of selected glasses to immobilize rare earth oxides waste

    Energy Technology Data Exchange (ETDEWEB)

    Mohd Fadzil, Syazwani, E-mail: mfsyazwani86@postech.ac.kr [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 790784 Pohang, Gyeongbuk (Korea, Republic of); School of Applied Physics, Faculty of Science and Technology, The National University of Malaysia, 43650 Bandar Baru Bangi, Selangor (Malaysia); Hrma, Pavel [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 790784 Pohang, Gyeongbuk (Korea, Republic of); Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA (United States); Schweiger, Michael J.; Riley, Brian J. [Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA (United States)

    2015-10-15

    Pyroprocessing is are processing method for managing and reusing used nuclear fuel (UNF) by dissolving it in an electrorefiner with a molten alkali or alkaline earth chloride salt mixture while avoiding wet reprocessing. Pyroprocessing UNF with a LiCl–KCl eutectic salt releases the fission products from the fuel and generates a variety of metallic and salt-based species, including rare earth (RE) chlorides. If the RE-chlorides are converted to oxides, borosilicate glass is a prime candidate for their immobilization because of its durability and ability to dissolve almost any RE waste component into the glass matrix at high loadings. Crystallization that occurs in waste glasses as the waste loading increases may complicate glass processing and affect the product quality. This work compares three types of borosilicate glasses in terms of liquidus temperature (T{sub L}): the International Simple Glass designed by the International Working Group, sodium borosilicate glass developed by Korea Hydro and Nuclear Power, and the lanthanide aluminoborosilicate (LABS) glass established in the United States. The LABS glass allows the highest waste loadings (over 50 mass% RE{sub 2}O{sub 3}) while possessing an acceptable chemical durability. - Highlights: • We investigated crystallization in borosilicate glasses containing rare earth oxides. • New crystallinity and durability data are shown for glasses proposed in the literature. • Both liquidus temperature and chemical durability increased as the waste loading increased.

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

    International Nuclear Information System (INIS)

    Plecas, I.; Dimovic, S.

    2006-01-01

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

  2. Ceramic process and plant design for high-level nuclear waste immobilization

    International Nuclear Information System (INIS)

    Grantham, L.F.; McKisson, R.L.; De Wames, R.E.; Guon, J.; Flintoff, J.F.; McKenzie, D.E.

    1983-01-01

    In the last 3 years, significant advances in ceramic technology for high-level nuclear waste solidification have been made. Product quality in terms of leach-resistance, compositional uniformity, structural integrity, and thermal stability promises to be superior to borosilicate glass. This paper addresses the process effectiveness and preliminary designs for glass and ceramic immobilization plants. The reference two-step ceramic process utilizes fluid-bed calcination (FBC) and hot isostatic press (HIP) consolidation. Full-scale demonstration of these well-developed processing steps has been established at DOE and/or commercial facilities for processing radioactive materials. Based on Savannah River-type waste, our model predicts that the capital and operating cost for the solidification of high-level nuclear waste is about the same for the ceramic and glass options. However, when repository costs are included, the ceramic option potentially offers significantly better economics due to its high waste loading and volume reduction. Volume reduction impacts several figures of merit in addition to cost such as system logistics, storage, transportation, and risk. The study concludes that the ceramic product/process has many potential advantages, and rapid deployment of the technology could be realized due to full-scale demonstrations of FBC and HIP technology in radioactive environments. Based on our finding and those of others, the ceramic innovation not only offers a viable backup to the glass reference process but promises to be a viable future option for new high-level nuclear waste management opportunities

  3. Development and characterization of basalt-glass ceramics for the immobilization of transuranic wastes

    International Nuclear Information System (INIS)

    Lokken, R.O.; Chick, L.A.; Thomas, L.E.

    1982-09-01

    Basalt-based waste forms were developed for the immobilization of transuranic (TRU) contaminated wastes. The specific waste studied is a 3:1 blend of process sludge and incinerator ash. Various amounts of TRU blended waste were melted with Pomona basalt powder. The vitreous products were subjected to a variety of heat treatment conditions to form glass ceramics. The total crystallinity of the glass ceramic, ranging from 20 to 45 wt %, was moderately dependent on composition and heat treatment conditions. Three parent glasses and four glass ceramics with varied composition and heat treatment were produced for detailed phase characterization and leaching. Both parent glasses and glass ceramics were mainly composed of a continuous, glassy matrix phase. This glass matrix entered into solution during leaching in both types of materials. The Fe-Ti rich dispersed glass phase was not significantly degraded by leaching. The glass ceramics, however, exhibited four to ten times less elemental releases during leaching than the parent glasses. The glass ceramic matrix probably contains higher Fe and Na and lower Ca and Mg relative to the parent glass matrix. The crystallization of augite in the glass ceramics is believed to contribute to the improved leach rates. Leach rates of the basalt glass ceramic are compared to those of other TRU nuclear waste forms containing 239 Pu

  4. Statement of work for the immobilized high-level waste transportation system, Project W-464

    Energy Technology Data Exchange (ETDEWEB)

    Mouette, P.

    1998-06-24

    The objective of this Statement of Work (SOW) is to present the scope, the deliverables, the organization, the technical and schedule expectations for the development of a Package Design Criteria (PDC), cost and schedule estimate for the acquisition of a transportation system for the Immobilized High-Level Waste (IHLW). This transportation system which includes the truck, the trailer, and a shielded cask will be used for on-site transportation of the IHLW canisters from the private vendor vitrification facility to the Hanford Site interim storage facility, i.e., vaults 2 and 3 of the Canister Storage Building (CSB). This Statement of Work asks Waste Management Federal Services, Inc., Northwest Operations, to provide Project W-464 with a Design Criteria Document, plus a life-cycle schedule and cost estimate for the acquisition of a transportation system (shielded cask, truck, trailer) for IHLW on-site transportation.

  5. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of tellurite glass as a waste form for salt wastes from electrochemical processing. The capacities to immobilize different salts were evaluated including: a LiCl-Li2O oxide reduction salt (for oxide fuel) containing fission products, a LiCl-KCl eutectic salt (for metallic fuel) containing fission products, and SrCl2. Physical and chemical properties of the glasses were characterized by using X-ray diffraction, bulk density measurements, chemical durability tests, scanning electron microscopy, and energy dispersive X-ray emission spectroscopy. These glasses were found to accommodate high concentrations of halide salts and have high densities. However, improvements are needed to meet chemical durability requirements.

  6. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

  7. Utilization of natural hematite as reactive barrier for immobilization of radionuclides from radioactive liquid waste.

    Science.gov (United States)

    El Afifi, E M; Attallah, M F; Borai, E H

    2016-01-01

    Potential utilization of hematite as a natural material for immobilization of long-lived radionuclides from radioactive liquid waste was investigated. Hematite ore has been characterized by different analytical tools such as Fourier transformer infrared (FTIR), X-ray fluorescence (XRF), powder X-ray diffraction (XRD), thermogravimetry (TG) and differential thermal (DT) analysis, scanning electron microscopy (SEM) and BET-surface area. In this study, europium was used as REEs(III) and as a homolog of Am(III)-isotopes (such as (241)Am of 432.6 y, (242m)Am of 141 y and (243)Am of 7370 y). Micro particles of the hematite ore were used for treatment of radioactive waste containing (152+154)Eu(III). The results indicated that 96% (4.1 × 10(4) Bq) of (152+154)Eu(III) was efficiently retained onto hematite ore. Kinetic experiments indicated that the processes could be simulated by a pseudo-second-order model and suggested that the process may be chemisorption in nature. The applicability of Langmuir, Freundlich and Temkin models was investigated. It was found that Langmuir isotherm exhibited the best fit with the experimental results. It can be concluded that hematite is an economic and efficient reactive barrier for immobilization of long-lived radio isotopes of actinides and REEs(III). Copyright © 2015 Elsevier Ltd. All rights reserved.

  8. Specialty glass development for radiation shielding windows and nuclear waste immobilization

    International Nuclear Information System (INIS)

    Mandal, S.; Ghorui, S.; Roy Chowdhury, A.; Sen, R.; Chakraborty, A.K.; Sen, S.; Maiti, H.S.

    2015-01-01

    The technology of two important varieties of specialty glasses, namely high density Radiation Shielding Window (RSW) glass and specialty glass beads of borosilicate composition have been successfully developed in CGCRI with an aim to meet the countries requirement. Radiation Shielding Windows used in nuclear installations, are viewing devices, which allow direct viewing into radioactive areas while still providing adequate protection to the operating personnel. The glass blocks are stabilized against damage from radiation by introducing cerium in definite proportions. Considering the essentially of developing an indigenous technology to make the country self-sufficient for this critical item, CGCRI has taken up a major programme to develop high lead containing glasses required for RSWs under a MoD with BARC. On the other hand, the specialty glass bead of specific composition and properties is a critical material required for management of radioactive waste in a closed nuclear fuel cycle that is followed by India. During reprocessing of the spent nuclear fuel, high level radio-active liquid waste (HLW) is produced containing unwanted radio isotopes some of which remain radioactive for thousands of years. The need is to immobilize them within a molecular structure so that they will not come out and be released to the ambience and thereby needs to be resolved if nuclear power is to make a significant contribution to the country's power requirement. Borosilicate glass has emerged as the material of choice for immobilization due to its unique random network structure

  9. Evaluation and review of alternative waste forms for immobilization of high level radioactive wastes

    International Nuclear Information System (INIS)

    1979-01-01

    Objective was to review the relative merits and potential of eleven alternative waste forms being considered for the solidification and disposal of radioactive wastes. A numerical rating of the alternative waste forms was arrived at individually by peer review panel members taking into consideration nine scientific and nine engineering parameters affecting the long-term performance and production of waste forms. A group rating for the alternative forms was achieved by averaging the individiual scores and discussing the available data base. Three final ranking lists comparing: (A) Present Scientific Merits or Least Risk for Use Today; (B) Research Priority; and (3) Present and Potential Engineering Practicality were prepared by the Panel. Each waste form in the lists is assigned a value of either (1) Top Rank, (2) Intermediate Rank, or is assigned a value of either (1) Top Rank, (2) Intermediate Rank, or (3) Bottom Rank. Relative strengths and weaknesses of the alternative waste forms and recommendations for future program directions are discussed

  10. Apatite and sodalite based glass-bonded waste forms for immobilization of 129I and mixed halide radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Goel, Ashutosh [Rutgers Univ., New Brunswick, NJ (United States); McCloy, John S. [Washington State Univ., Pullman, WA (United States); Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-12-30

    The goal of the project was to utilize the knowledge accumulated by the team, in working with minerals for chloride wastes and biological apatites, toward the development of advanced waste forms for immobilizing 129I and mixed-halide wastes. Based on our knowledge, experience, and thorough literature review, we had selected two minerals with different crystal structures and potential for high chemical durability, sodalite and CaP/PbV-apatite, to form the basis of this project. The focus of the proposed effort was towards: (i) low temperature synthesis of proposed minerals (iodine containing sodalite and apatite) leading to the development of monolithic waste forms, (ii) development of a fundamental understanding of the atomic-scale to meso-scale mechanisms of radionuclide incorporation in them, and (iii) understanding of the mechanism of their chemical corrosion, alteration mechanism, and rates. The proposed work was divided into four broad sections. deliverables. 1. Synthesis of materials 2. Materials structural and thermal characterization 3. Design of glass compositions and synthesis glass-bonded minerals, and 4. Chemical durability testing of materials.

  11. Standard test method for determining liquidus temperature of immobilized waste glasses and simulated waste glasses

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 These practices cover procedures for determining the liquidus temperature (TL) of nuclear waste, mixed nuclear waste, simulated nuclear waste, or hazardous waste glass in the temperature range from 600°C to 1600°C. This method differs from Practice C829 in that it employs additional methods to determine TL. TL is useful in waste glass plant operation, glass formulation, and melter design to determine the minimum temperature that must be maintained in a waste glass melt to make sure that crystallization does not occur or is below a particular constraint, for example, 1 volume % crystallinity or T1%. As of now, many institutions studying waste and simulated waste vitrification are not in agreement regarding this constraint (1). 1.2 Three methods are included, differing in (1) the type of equipment available to the analyst (that is, type of furnace and characterization equipment), (2) the quantity of glass available to the analyst, (3) the precision and accuracy desired for the measurement, and (4) candi...

  12. Recent Improvements In Interface Management For Hanfords Waste Treatment And Immobilization Plant - 13263

    International Nuclear Information System (INIS)

    Arm, Stuart T.; Pell, Michael J.; Van Meighem, Jeffery S.; Duncan, Garth M.; Harrington, Christopher C.

    2012-01-01

    The U.S. Department of Energy (DOE), Office of River Protection (ORP) is responsible for management and completion of the River Protection Project (RPP) mission, which comprises both the Hanford Site tank farms operations and the Waste Treatment and Immobilization Plant (WTP). The RPP mission is to store, retrieve and treat Hanford's tank waste; store and dispose of treated wastes; and close the tank farm waste management areas and treatment facilities by 2047. The WTP is currently being designed and constructed by Bechtel National Inc. (BNI) for DOE-ORP. BNI relies on a number of technical services from other Hanford contractors for WTP's construction and commissioning. These same services will be required of the future WTP operations contractor. The WTP interface management process has recently been improved through changes in organization and technical issue management documented in an Interface Management Plan. Ten of the thirteen active WTP Interface Control Documents (ICDs) have been revised in 2012 using the improved process with the remaining three in progress. The value of the process improvements is reflected by the ability to issue these documents on schedule

  13. US DOE Initiated Performance Enhancements to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-activity Waste Vitrification (LAW) System

    International Nuclear Information System (INIS)

    Hamel, William F.; Gerdes, Kurt D.; Holton, Langdon K.; Pegg, Ian L.; Bowen, Brad W.

    2006-01-01

    The U.S Department of Energy Office of River Protection (DOE-ORP) is constructing a Waste Treatment and Immobilization Plant (WTP) for the treatment and vitrification of underground tank wastes stored at the Hanford Site in Washington State. The WTP comprises four major facilities: a pretreatment facility to separate the tank waste into high level waste (HLW) and low-activity waste (LAW) process streams, a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction, and an analytical laboratory to support the operations of all four treatment facilities. DOE has established strategic objectives to optimize the performance of the WTP facilities and the LAW and HLW waste forms to reduce the overall schedule and cost for treatment and vitrification of the Hanford tank wastes. This strategy has been implemented by establishing performance expectations in the WTP contract for the facilities and waste forms. In addition, DOE, as owner-operator of the WTP facilities, continues to evaluate (1) the design, to determine the potential for performance above the requirements specified in the WTP contract; and (2) improvements in production of the LAW and HLW waste forms. This paper reports recent progress directed at improving production of the LAW waste form. DOE's initial assessment, which is based on the work reported in this paper, is that the capacity of the WTP LAW vitrification facility can be increased by a factor of 2 to 4 with a combination of revised glass formulations, modest increases in melter glass operating temperatures, and a second-generation LAW melter with a larger surface area. Implementing these improvements in the LAW waste immobilization capability can benefit the LAW treatment mission by reducing both processing time and cost

  14. Setting and stiffening of cementitious components in Cast Stone waste form for disposal of secondary wastes from the Hanford waste treatment and immobilization plant

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chul-Woo; Chun, Jaehun, E-mail: jaehun.chun@pnnl.gov; Um, Wooyong; Sundaram, S.K.; Westsik, Joseph H.

    2013-04-01

    Cast Stone is a cementitious waste form, a viable option to immobilize secondary nuclear liquid wastes generated from the Hanford Waste Treatment and Immobilization Plant. However, no study has been performed to understand the flow and stiffening behavior, which is essential to ensure proper workability and is important to safety in a nuclear waste field-scale application. X-ray diffraction, rheology, and ultrasonic wave reflection methods were used to understand the specific phase formation and stiffening of Cast Stone. Our results showed a good correlation between rheological properties of the fresh mixture and phase formation in Cast Stone. Secondary gypsum formation was observed with low concentration simulants, and the formation of gypsum was suppressed in high concentration simulants. A threshold concentration for the drastic change in stiffening was found at 1.56 M Na concentration. It was found that the stiffening of Cast Stone was strongly dependent on the concentration of simulant. Highlights: • A combination of XRD, UWR, and rheology gives a better understanding of Cast Stone. • Stiffening of Cast Stone was strongly dependent on the concentration of simulant. • A drastic change in stiffening of Cast Stone was found at 1.56 M Na concentration.

  15. Evaluation and review of alternative waste forms for immobilization of high level radioactive wastes

    International Nuclear Information System (INIS)

    1980-01-01

    The objective of this study was to review the relative merits and potential of 15 (fifteen) alternative waste forms being considered for the solidification and disposal of radioactive wastes. The relative merits of 4 (four) alternative pre-solidification processing approaches were also assessed in this study. A Peer Review Panel composed of 8 (eight) scientists and engineers representing independent, non-DOE laboratories from industry, government, and universities and the disciplines of materials science, ceramics, glass, metallurgy, and geology conducted the review. A numerical rating of alternative waste forms was arrived at individually by the panel members taking into consideration 9 (nine) scientific and 9 (nine) engineering parameters affecting the long term performance and production of waste forms. At a meeting on May 9, 1980, a group ranking for the alternative forms was achieved by averaging the individual scores and discussing the available data base. Three final ranking lists comparing: (A) Present Scientific Merits or Least Risk for Use Today; and (B) Research Priority; and (C) Present and Potential Engineering Practicality were prepared by the Panel. Each waste form in the lists is assigned a value of either (1) Top Rank, (2) Intermediate Rank, or (3) Bottom Rank. A discussion of the relative strengths and weaknesses of the alternative waste forms and recommendations for future program directions is presented in the body of the accompanying Peer Review Panel report

  16. Immobilized High-Level Waste (HLW) Interim Storage Alternative Generation and analysis and Decision Report - second Generation Implementing Architecture

    International Nuclear Information System (INIS)

    CALMUS, R.B.

    2000-01-01

    Two alternative approaches were previously identified to provide second-generation interim storage of Immobilized High-Level Waste (IHLW). One approach was retrofit modification of the Fuel and Materials Examination Facility (FMEF) to accommodate IHLW. The results of the evaluation of the FMEF as the second-generation IHLW interim storage facility and subsequent decision process are provided in this document

  17. Packaging design criteria (onsite) project W-520 immobilized low-activity waste transportation system

    International Nuclear Information System (INIS)

    BOEHNKE, W.M.

    2001-01-01

    A plan is currently in place to process the high-level radioactive wastes that resulted from uranium and plutonium recovery operations from Spent Nuclear Fuel at the Hanford Site, Richland, Washington. Currently, millions of gallons of high-level radioactive waste in the form of liquids, sludges, and saltcake are stored in many large underground tanks onsite. This waste will be processed and separated into high-level and low-activity fractions. Both fractions will then be vitrified (i.e., blended with molten borosilicate glass) in order to encapsulate the toxic radionuclides. The immobilized low-activity waste (ILAW) glass will be poured into LAW canisters, allowed to cool and harden to solid form, sealed by welding, and then transported to a double-lined trench in the 200 East Area for permanent disposal. This document presents the packaging design criteria (PDC) for an onsite LAW transportation system, which includes the ILAW canister, ILAW package, and transport vehicle and defines normal and accident conditions. This PDC provides the basis for the ILAW onsite transportation system design and fabrication and establishes the transportation safety criteria that the design will be evaluated against in the Package Specific Safety Document (PSSD). It provides the criteria for the ILAW canister, cask and transport vehicles and defines normal and accident conditions. The LAW transportation system is designed to transport stabilized waste from the vitrification facility to the ILAW disposal facility developed by Project W-520. All ILAW transport will take place within the 200 East Area (all within the Hanford Site)

  18. Immobilization of high level nuclear wastes in sintered glasses. Devitrification evaluation produced with different thermal treatments

    International Nuclear Information System (INIS)

    Messi de Bernasconi, N.B.; Russo, D.O.; Bevilacqua, M.E.; Sterba, M.E.; Heredia, A.D.; Audero, M.A.

    1990-01-01

    This work describes immobilization of high level nuclear wastes in sintered glass, as alternative way to melting glass. Different chemical compositions of borosilicate glass with simulate waste were utilized and satisfactory results were obtained at laboratory scale. As another contribution to the materials studies by X ray powder diffraction analysis, the devitrification produced with different thermal treatments, was evaluated. The effect of the thermal history on the behaviour of fission products containing glasses has been studied by several working groups in the field of high level waste fixation. When the glass is cooled through the temperature range from 800 deg C down to less than 400 deg C (these temperatures are approximates) nucleation and crystal growth can take place. The rate of crystallization will be maximum near the transformation point but through this rate may be low at lower temperatures, devitrification can still occur over long periods of time, depending on the glass composition. It was verified that there can be an appreciable increase in leaching in some waste glass compositions owing to the presence of crystalline phases. On the other hand, other compositions show very little change in leachability and the devitrified product is often preferable as there is less tendency to cracking, particularly in massive blocks of glass. A borosilicate glass, named SG7, which was developed specially in the KfK for the hot pressing of HLW with glass frit was studied. It presents a much enhanced chemical durability than borosolicate glass developed for the melting process. The crystallization behaviour of SG7 glass products was investigated in our own experiments by annealing sintered samples up to 3000 h at temperatures between 675 and 825 deg C. The samples had contained simulated waste with noble metals, since these might act as foreign nuclei for crystallization. Results on the extent of devitrification and time- temperature- transformation curves are

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

    International Nuclear Information System (INIS)

    Giraldelli, M.A.

    1990-01-01

    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)

  20. Synthesis of biodiesel from waste cooking oil using immobilized lipase in fixed bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen Yingming [School of Environment and Urban Construction, Wuhan University of Science and Engineering, Wuhan 430073 (China); Guangzhou Institute of Energy Conversion, Chinese Academy of Science, Guangzhou 510640 (China); Xiao Bo [School of Environmental Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Chang Jie [School of Chemistry and Chemical Engineering, South China University of Technology, Guangzhou 510641 (China)], E-mail: changjie@scut.edu.cn; Fu Yan [School of Chemistry and Chemical Engineering, South China University of Technology, Guangzhou 510641 (China); Lv Pengmei; Wang Xuewei [Guangzhou Institute of Energy Conversion, Chinese Academy of Science, Guangzhou 510640 (China)

    2009-03-15

    Waste cooking oil (WCO) is the residue from the kitchen, restaurants, food factories and even human and animal waste which not only harm people's health but also causes environmental pollution. The production of biodiesel from waste cooking oil to partially substitute petroleum diesel is one of the measures for solving the twin problems of environment pollution and energy shortage. In this project, synthesis of biodiesel was catalyzed by immobilized Candida lipase in a three-step fixed bed reactor. The reaction solution was a mixture of WCO, water, methanol and solvent (hexane). The main product was biodiesel consisted of fatty acid methyl ester (FAME), of which methyl oleate was the main component. Effects of lipase, solvent, water, and temperature and flow of the reaction mixture on the synthesis of biodiesel were analyzed. The results indicate that a 91.08% of FAME can be achieved in the end product under optimum conditions. Most of the chemical and physical characters of the biodiesel were superior to the standards for 0 diesel (GB/T 19147) and biodiesel (DIN V51606 and ASTM D-6751)

  1. Synthesis of biodiesel from waste cooking oil using immobilized lipase in fixed bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yingming [School of Environment and Urban Construction, Wuhan University of Science and Engineering, Wuhan 430073 (China)]|[Guangzhou Institute of Energy Conversion, Chinese Academy of Science, Guangzhou 510640 (China); Xiao, Bo [School of Environmental Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Chang, Jie; Fu, Yan [School of Chemistry and Chemical Engineering, South China University of Technology, Guangzhou 510641 (China); Lv, Pengmei; Wang, Xuewei [Guangzhou Institute of Energy Conversion, Chinese Academy of Science, Guangzhou 510640 (China)

    2009-03-15

    Waste cooking oil (WCO) is the residue from the kitchen, restaurants, food factories and even human and animal waste which not only harm people's health but also causes environmental pollution. The production of biodiesel from waste cooking oil to partially substitute petroleum diesel is one of the measures for solving the twin problems of environment pollution and energy shortage. In this project, synthesis of biodiesel was catalyzed by immobilized Candida lipase in a three-step fixed bed reactor. The reaction solution was a mixture of WCO, water, methanol and solvent (hexane). The main product was biodiesel consisted of fatty acid methyl ester (FAME), of which methyl oleate was the main component. Effects of lipase, solvent, water, and temperature and flow of the reaction mixture on the synthesis of biodiesel were analyzed. The results indicate that a 91.08% of FAME can be achieved in the end product under optimum conditions. Most of the chemical and physical characters of the biodiesel were superior to the standards for 0diesel (GB/T 19147) and biodiesel (DIN V51606 and ASTM D-6751). (author)

  2. Synthesis of biodiesel from waste cooking oil using immobilized lipase in fixed bed reactor

    International Nuclear Information System (INIS)

    Chen Yingming; Xiao Bo; Chang Jie; Fu Yan; Lv Pengmei; Wang Xuewei

    2009-01-01

    Waste cooking oil (WCO) is the residue from the kitchen, restaurants, food factories and even human and animal waste which not only harm people's health but also causes environmental pollution. The production of biodiesel from waste cooking oil to partially substitute petroleum diesel is one of the measures for solving the twin problems of environment pollution and energy shortage. In this project, synthesis of biodiesel was catalyzed by immobilized Candida lipase in a three-step fixed bed reactor. The reaction solution was a mixture of WCO, water, methanol and solvent (hexane). The main product was biodiesel consisted of fatty acid methyl ester (FAME), of which methyl oleate was the main component. Effects of lipase, solvent, water, and temperature and flow of the reaction mixture on the synthesis of biodiesel were analyzed. The results indicate that a 91.08% of FAME can be achieved in the end product under optimum conditions. Most of the chemical and physical characters of the biodiesel were superior to the standards for 0 diesel (GB/T 19147) and biodiesel (DIN V51606 and ASTM D-6751)

  3. TWRS Retrieval and Storage Mission and Immobilized Low Activity Waste (ILAW) Disposal Plan

    International Nuclear Information System (INIS)

    BURBANK, D.A.

    1999-01-01

    This project plan has a twofold purpose. First, it provides a waste stream project plan specific to the River Protection Project (RPP) (formerly the Tank Waste Remediation System [TWRS] Project) Immobilized Low-Activity Waste (LAW) Disposal Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-90-01 (Ecology et al. 1994) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan (Ecology et al. 1998). Second, it provides an upper tier document that can be used as the basis for future subproject line-item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 [DOE 1992] and 430.1 [DOE 1995a]). The format and content of this project plan are designed to accommodate the requirements mentioned by the Tri-Party Agreement and the DOE orders. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed

  4. Rhenium solubility in borosilicate nuclear waste glass: implications for the processing and immobilization of technetium-99.

    Science.gov (United States)

    McCloy, John S; Riley, Brian J; Goel, Ashutosh; Liezers, Martin; Schweiger, Michael J; Rodriguez, Carmen P; Hrma, Pavel; Kim, Dong-Sang; Lukens, Wayne W; Kruger, Albert A

    2012-11-20

    The immobilization of technetium-99 ((99)Tc) in a suitable host matrix has proven to be a challenging task for researchers in the nuclear waste community around the world. In this context, the present work reports on the solubility and retention of rhenium, a nonradioactive surrogate for (99)Tc, in a sodium borosilicate glass. Glasses containing target Re concentrations from 0 to 10,000 ppm [by mass, added as KReO(4) (Re(7+))] were synthesized in vacuum-sealed quartz ampules to minimize the loss of Re from volatilization during melting at 1000 °C. The rhenium was found as Re(7+) in all of the glasses as observed by X-ray absorption near-edge structure. The solubility of Re in borosilicate glasses was determined to be ~3000 ppm (by mass) using inductively coupled plasma optical emission spectroscopy. At higher rhenium concentrations, additional rhenium was retained in the glasses as crystalline inclusions of alkali perrhenates detected with X-ray diffraction. Since (99)Tc concentrations in a glass waste form are predicted to be wastes, assuming Tc as Tc(7+) and similarities between Re(7+) and Tc(7+) behavior in this glass system.

  5. Dark fermentative hydrogen production by defined mixed microbial cultures immobilized on ligno-cellulosic waste materials

    Energy Technology Data Exchange (ETDEWEB)

    Patel, Sanjay K.S. [Microbial Biotechnology and Genomics, Institute of Genomics and Integrative Biology (IGIB), CSIR, Delhi University Campus, Mall Road, Delhi 110007 (India); Department of Biotechnology, University of Pune, Pune 411007 (India); Purohit, Hemant J. [Environmental Genomics Unit, National Environmental Engineering Research Institute (NEERI), CSIR, Nehru Marg, Nagpur 440020 (India); Kalia, Vipin C. [Microbial Biotechnology and Genomics, Institute of Genomics and Integrative Biology (IGIB), CSIR, Delhi University Campus, Mall Road, Delhi 110007 (India)

    2010-10-15

    Mixed microbial cultures (MMCs) based on 11 isolates belonging to Bacillus spp. (Firmicutes), Bordetella avium, Enterobacter aerogenes and Proteus mirabilis (Proteobacteria) were employed to produce hydrogen (H{sub 2}) under dark fermentative conditions. Under daily fed culture conditions (hydraulic retention time of 2 days), MMC6 and MMC4, immobilized on ligno-cellulosic wastes - banana leaves and coconut coir evolved 300-330 mL H{sub 2}/day. Here, H{sub 2} constituted 58-62% of the total biogas evolved. It amounted to a H{sub 2} yield of 1.54-1.65 mol/mol glucose utilized over a period of 60 days of fermentation. The involvement of various Bacillus spp. -Bacillus sp., Bacillus cereus, Bacillus megaterium, Bacillus pumilus and Bacillus thuringiensis as components of the defined MMCs for H{sub 2} production has been reported here for the first time. (author)

  6. Geologic Data Package for 2001 Immobilized Low-Activity Waste Performance Assessment

    International Nuclear Information System (INIS)

    SP Reidel; DG Horton

    1999-01-01

    This database is a compilation of existing geologic data from both the existing and new immobilized low-activity waste disposal sites for use in the 2001 Performance Assessment. Data were compiled from both surface and subsurface geologic sources. Large-scale surface geologic maps, previously published, cover the entire 200-East Area and the disposal sites. Subsurface information consists of drilling and geophysical logs from nearby boreholes and stored sediment samples. Numerous published geological reports are available that describe the subsurface geology of the area. Site-specific subsurface data are summarized in tables and profiles in this document. Uncertainty in data is mainly restricted to borehole information. Variations in sampling and drilling techniques present some correlation uncertainties across the sites. A greater degree of uncertainty exists on the new site because of restricted borehole coverage. There is some uncertainty to the location and orientation of elastic dikes across the sites

  7. PRELIMINARY STUDY OF CERAMICS FOR IMMOBILIZATION OF ADVANCED FUEL CYCLE REPROCESSING WASTES

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-22

    The Savannah River National Laboratory (SRNL) developed a series of ceramic waste forms for the immobilization of Cesium/Lanthanide (CS/LN) and Cesium/Lanthanide/Transition Metal (CS/LN/TM) waste streams anticipated to result from nuclear fuel reprocessing. Simple raw materials, including Al{sub 2}O{sub 3}, CaO, and TiO{sub 2} were combined with simulated waste components to produce multiphase ceramics containing hollandite-type phases, perovskites (particularly BaTiO{sub 3}), pyrochlores, zirconolite, and other minor metal titanate phases. Identification of excess Al{sub 2}O{sub 3} via X-ray Diffraction (XRD) and Scanning Electron Microscopy with Energy Dispersive Spectroscopy (SEM/EDS) in the first series of compositions led to a Phase II study, with significantly reduced Al{sub 2}O{sub 3} concentrations and increased waste loadings. Three fabrication methodologies were used, including melting and crystallizing, pressing and sintering, and Spark Plasma Sintering (SPS), with the intent of studying phase evolution under various sintering conditions. XRD and SEM/EDS results showed that the partitioning of the waste elements in the sintered materials was very similar, despite varying stoichiometry of the phases formed. The Phase II compositions generally contained a reduced amount of unreacted Al{sub 2}O{sub 3} as identified by XRD, and had phase assemblages that were closer to the initial targets. Chemical composition measurements showed no significant issues with meeting the target compositions. However, volatilization of Cs and Mo was identified, particularly during melting, since sintering of the pressed pellets and SPS were performed at lower temperatures. Partitioning of some of the waste components was difficult to determine via XRD. SEM/EDS mapping showed that those elements, which were generally present in small concentrations, were well distributed throughout the waste forms. Initial studies of radiation damage tolerance using ion beam irradiation at Los

  8. Heavy metal immobilization in soil near abandoned mines using eggshell waste and rapeseed residue.

    Science.gov (United States)

    Lee, Sang Soo; Lim, Jung Eun; El-Azeem, Samy A M Abd; Choi, Bongsu; Oh, Sang-Eun; Moon, Deok Hyun; Ok, Yong Sik

    2013-03-01

    Heavy metal contamination of agricultural soils has received great concern due to potential risk to human health. Cadmium and Pb are largely released from abandoned or closed mines in Korea, resulting in soil contamination. The objective of this study was to evaluate the effects of eggshell waste in combination with the conventional nitrogen, phosphorous, and potassium fertilizer (also known as NPK fertilizer) or the rapeseed residue on immobilization of Cd and Pb in the rice paddy soil. Cadmium and Pb extractabilities were tested using two methods of (1) the toxicity characteristics leaching procedure (TCLP) and (2) the 0.1 M HCl extraction. With 5 % eggshell addition, the values of soil pH were increased from 6.33 and 6.51 to 8.15 and 8.04 in combination with NPK fertilizer and rapeseed residue, respectively, compared to no eggshell addition. The increase in soil pH may contribute to heavy metal immobilization by altering heavy metals into more stable in soils. Concentrations of TCLP-extracted Cd and Pb were reduced by up to 67.9 and 93.2 % by addition of 5 % eggshell compared to control. For 0.1 M HCl extraction method, the concentration of 0.1 M HCl-Cd in soils treated with NPK fertilizer and rapeseed residue was significantly reduced by up to 34.01 and 46.1 %, respectively, with 5 % eggshell addition compared to control. A decrease in acid phosphatase activity and an increase in alkaline phosphatase activity at high soil pH were also observed. Combined application of eggshell waste and rapeseed residue can be cost-effective and beneficial way to remediate the soil contaminated with heavy metals.

  9. Assessment of low density polyethylene characteristics for hazardous waste immobilization; Avaliacao das caracteristicas do polietileno de baixa densidade visando a imobilizacao de rejeitos perigosos

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Stela; Carvalho, Larissa Lara de; Pacheco, Graziella Rajao Cota; Oliveira, Tania Valeria de; Senne Junior, Murillo; Pacheco, Raquel R. Janot [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)]. E-mail: sdsc@cdtn.br

    2005-07-01

    This paper analyses the properties of low density polyethylene (LDPE) to allow choosing the more suitable to be used as matrices for radioactive and hazardous waste immobilization. Four virgin and recycled LDPE, with different melting index, were evaluated by extrusion and compressive strength tests. A preliminary immobilization test has been carried out using a simulated waste and one of the evaluated polymers and the homogeneity of the final waste product was determinate by the analysis of the material density. (author)

  10. External Criticality Risk of Immobilized Plutonium Waste Form in a Geologic Repository

    International Nuclear Information System (INIS)

    McClure, J.

    2001-01-01

    This purpose of this technical report is to provide a comprehensive summary of the waste package (WP) external criticality-related risk of the Plutonium Disposition ceramic waste form, which is being developed and evaluated by the Office of Fissile Materials Disposition of the United States Department of Energy (DOE). Potential accumulation of the fissile materials, 239 Pu and 235 U, in rock formations having a favorable chemical environment for such actions, requires analysis because autocatalytic configurations, while unlikely to form, never-the-less have consequences which are undesirable and require evaluation. Secondly, the WP design has evolved necessitating a re-evaluation of the internal WP degradation scenarios that contribute to the external source terms. The scope of this study includes a summary of the revised WP degradation calculations, a summary of the accumulation mechanisms in fractures and lithophysae in the tuff beneath the WP footprint, and a summary of the criticality risk calculations from any accumulated fissile material. Accumulations of fissile material external to the WP sufficient to pose a potential criticality risk require a deposition mechanism operating over sufficient time to reach required levels. The transporting solution concentrations themselves are well below critical levels (CRWMS 2001e). The ceramic waste form consists of Pu immobilized in ceramic disks, which would be embedded in High-Level Waste (HLW) glass in the standard HLW glass disposal canister. The ceramic disks would occupy approximately 12% of the HLW canister volume, while most of the remaining 88% of the volume would be occupied by HLW glass

  11. Near-Field Hydrology Data Package for the Immobilized Low-Activity Waste 2001 Performance Assessment

    International Nuclear Information System (INIS)

    PD Meyer; RJ Serne

    1999-01-01

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are currently stored in single- and double-shell tanks at the Hanford Site. The preferred method for disposing of the portion that is classified as immobilized low-activity waste (ILAW) is to vitrify the waste and place the product in new-surface, shallow land burial facilities. The LMHC project to assess the performance of these disposal facilities is the Hanford ILAW Performance Assessment (PA) Activity. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface water resources, and inadvertent intruders. Achieving this goal will require prediction of contaminant migration from the facilities. This migration is expected to occur primarily via the movement of water through the facilities and the consequent transport of dissolved contaminants in the pore water of the vadose zone. Pacific Northwest National Laboratory (PNNL) assists LMHC in its performance assessment activities. One of PNNL's tasks is to provide estimates of the physical, hydraulic, and transport properties of the materials comprising the disposal facilities and the disturbed region around them. These materials are referred to as the near-field materials. Their properties are expressed as parameters of constitutive models used in simulations of subsurface flow and transport. In addition to the best-estimate parameter values, information on uncertainty in the parameter values and estimates of the changes in parameter values over time are required to complete the PA. These parameter estimates and information are contained in this report, the Near-Field Hydrology Data Package

  12. Recent Improvements in Interface Management for Hanford's Waste Treatment and Immobilization Plant - 13263

    Energy Technology Data Exchange (ETDEWEB)

    Arm, Stuart T.; Van Meighem, Jeffery S. [Washington River Protection Solutions, P.O. Box 850, Richland, Washington, 99352 (United States); Duncan, Garth M.; Pell, Michael J. [Bechtel National Inc., 2435 Stevens Center Place, Richland, Washington, 99352 (United States); Harrington, Christopher C. [Department of Energy - Office of River Protection, 2440 Stevens Center Place, Richland, Washington, 99352 (United States)

    2013-07-01

    The U.S. Department of Energy (DOE), Office of River Protection (ORP) is responsible for management and completion of the River Protection Project (RPP) mission, which includes the Hanford Site tank farms operations and the Waste Treatment and Immobilization Plant (WTP). The RPP mission is to store, retrieve and treat Hanford's tank waste; store and dispose of treated wastes; and close the tank farm waste management areas and treatment facilities by 2047. The WTP is currently being designed and constructed by Bechtel National Inc. (BNI) for DOE-ORP. BNI relies on a number of technical services from other Hanford contractors for WTP's construction and commissioning. These same services will be required of the future WTP operations contractor. Partly in response to a DNFSB recommendation, the WTP interface management process managing these technical services has recently been improved through changes in organization and issue management. The changes are documented in an Interface Management Plan. The organizational improvement is embodied in the One System Integrated Project Team that was formed by integrating WTP and tank farms staff representing interfacing functional areas into a single organization. A number of improvements were made to the issue management process but most notable was the formal appointment of technical, regulatory and safety subject matter experts to ensure accurate identification of issues and open items. Ten of the thirteen active WTP Interface Control Documents have been revised in 2012 using the improved process with the remaining three in progress. The value of the process improvements is reflected by the ability to issue these documents on schedule and accurately identify technical, regulatory and safety issues and open items. (authors)

  13. Optimized production of biodiesel from waste cooking oil by lipase immobilized on magnetic nanoparticles.

    Science.gov (United States)

    Yu, Chi-Yang; Huang, Liang-Yu; Kuan, I-Ching; Lee, Shiow-Ling

    2013-12-11

    Biodiesel, a non-toxic and biodegradable fuel, has recently become a major source of renewable alternative fuels. Utilization of lipase as a biocatalyst to produce biodiesel has advantages over common alkaline catalysts such as mild reaction conditions, easy product separation, and use of waste cooking oil as raw material. In this study, Pseudomonas cepacia lipase immobilized onto magnetic nanoparticles (MNP) was used for biodiesel production from waste cooking oil. The optimal dosage of lipase-bound MNP was 40% (w/w of oil) and there was little difference between stepwise addition of methanol at 12 h- and 24 h-intervals. Reaction temperature, substrate molar ratio (methanol/oil), and water content (w/w of oil) were optimized using response surface methodology (RSM). The optimal reaction conditions were 44.2 °C, substrate molar ratio of 5.2, and water content of 12.5%. The predicted and experimental molar conversions of fatty acid methyl esters (FAME) were 80% and 79%, respectively.

  14. Production and immobilization of enzymes by solid-state fermentation of agroindustrial waste.

    Science.gov (United States)

    Romo Sánchez, Sheila; Gil Sánchez, Irene; Arévalo-Villena, María; Briones Pérez, Ana

    2015-03-01

    The recovery of by-products from agri-food industry is currently one of the major challenges of biotechnology. Castilla-La Mancha produces around three million tons of waste coming from olive oil and wine industries, both of which have a pivotal role in the economy of this region. For this reason, this study reports on the exploitation of grape skins and olive pomaces for the production of lignocellulosic enzymes, which are able to deconstruct the agroindustrial waste and, therefore, reuse them in future industrial processes. To this end, solid-state fermentation was carried out using two local fungal strains (Aspergillus niger-113 N and Aspergillus fumigatus-3). In some trials, a wheat supplementation with a 1:1 ratio was used to improve the growth conditions, and the particle size of the substrates was altered through milling. Separate fermentations were run and collected after 2, 4, 6, 8, 10 and 15 days to monitor enzymatic activity (xylanase, cellulase, β-glucosidase, pectinase). The highest values were recorded after 10 and 15 days of fermentation. The use of A. niger on unmilled grape skin yielded the best outcomes (47.05 U xylanase/g by-product). The multi-enzymatic extracts obtained were purified, freeze dried, and immobilized on chitosan by adsorption to assess the possible advantages provided by the different techniques.

  15. TWRS retrieval and disposal mission. Immobilized high-level waste storage plan

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1998-01-01

    This project plan has a two fold purpose. First, it provides a plan specific to the Hanford Tank Waste Remediation System (TWRS) Immobilized High-Level Waste (EMW) Storage Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) milestone M-90-01 (Ecology et al. 1996) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan. Second, it provides an upper tier document that can be used as the basis for future subproject line item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 (DOE 1992a) and 430.1 (DOE 1995)). The format and content of this project plan are designed to accommodate the plan's dual purpose. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed

  16. Testing and evaluation of the properties of various potential materials for immobilizing high activity waste

    International Nuclear Information System (INIS)

    Malow, G.; Beran, V.; Lutze, W.

    1980-01-01

    Second joint annual report of the work performed on the testing and evaluation of materials for immobilizing high activity waste under Community contracts. At Marcoule, active block samples containing HAW from the Marcoule reprocessing plant were cast to the specification of five of the six original reference samples and leach tested at ambient temperature. Phosphate glass -bead samples produced by the Gelsenberg/DWK PAMELA process- were included in the test programme at HMI-Berlin and UKAEA Harwell. Leaching tests of inactive samples with leachants of various pH-values, with ionized water and with natural water compositions representative of certain repository conditions (salt, clay and granite) were added to the Harwell programme. The studies of radiation and thermal effects and the investigation of devitrification phenomena, started in 1977, continued, as samples reached annealing times of 2400 h and doses 4 x 10 17 dpg. The results achieved have so far confirmed most of the favourable preliminary assessments of glass based solidification products. At this stage the programme aims primarily at the understanding of physical and chemical phenomena rather that at verification under realistic waste storage and disposal conditions

  17. Radiation effects in glass and glass-ceramic waste forms for the immobilization of CANDU UO{sub 2} fuel reprocessing waste

    Energy Technology Data Exchange (ETDEWEB)

    Tait, J C

    1993-05-01

    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 {sup 129}I, {sup 85}Kr and {sup 14}C. (author). 104 refs., 9 tabs., 5 figs.

  18. Glass-bonded iodosodalite waste form for immobilization of 129I

    Science.gov (United States)

    Chong, Saehwa; Peterson, Jacob A.; Riley, Brian J.; Tabada, Diana; Wall, Donald; Corkhill, Claire L.; McCloy, John S.

    2018-06-01

    Immobilization of radioiodine is an important requirement for current and future nuclear fuel cycles. Iodosodalite [Na8(AlSiO4)6I2] was synthesized hydrothermally from metakaolin, NaI, and NaOH. Dried unwashed sodalite powders were used to synthesize glass-bonded iodosodalite waste forms (glass composite materials) by heating pressed pellets at 650, 750, or 850 °C with two types of sodium borosilicate glass binders. These heat-treated specimens were characterized with X-ray diffraction, Fourier-transform infrared spectroscopy, scanning electron microscopy, energy dispersive spectroscopy, thermal analysis, porosity and density measurements, neutron activation analysis, and inductively-coupled plasma mass spectrometry. For the best waste form produced (pellets mixed with 10 mass% of glass binder and heat-treated at 750 °C), the maximum possible elemental iodine loading was 19.8 mass%, but only ∼8-9 mass% waste loading of iodine was retained in the waste form after thermal processing. Other pellets with higher iodine retention either contained higher porosity or were incompletely sintered. ASTM C1308 and C1285 (product consistency test, PCT) experiments were performed to understand chemical durability under diffusive and static conditions. The C1308 test resulted in significantly higher normalized loss compared to the C1285 test, most likely because of the strong effect of neutral pH solution renewal and prevention of ion saturation in solution. Both experiments indicated that release rates of Na and Si were higher than for Al and I, probably due to a poorly durable Na-Si-O phase from the glass bonding matrix or from initial sodalite synthesis; however the C1308 test result indicated that congruent dissolution of iodosodalite occurred. The average release rates of iodine obtained from C1308 were 0.17 and 1.29 g m-2 d-1 for 80 or 8 m-1, respectively, and the C1285 analysis gave a value of 2 × 10-5 g m-2 d-1, which is comparable to or better than the durability of

  19. Effect of immobilized biosorbents on the heavy metals (Cu2+) biosorption with variations of temperature and initial concentration of waste

    Science.gov (United States)

    Siwi, W. P.; Rinanti, A.; Silalahi, M. D. S.; Hadisoebroto, R.; Fachrul, M. F.

    2018-01-01

    The aims of research is to studying the efficiency of copper removal by combining immobilized microalgae with optimizations of temperature and initial Copper concentration. The research was conducted in batch culture with temperature variations of 25°C, 30°C, and 35°C, as well as initial Cu2+ concentrations (mg/l) of 3, 5, 10, 15 and 20 using monoculture of S. cerevisiae, Chlorella sp., and mixed culture of them both as immobilized biosorbents. The optimum adsorption of 83.4% obtained in temperature of 30°C with an initial waste concentration of 17.62 mg/l, initial biomass concentration of 200 mg, pH of 4, and 120 minutes detention time by the immobilized mixed culture biosorbent. The cell morphology examined using Scanning Electron Microscope (SEM) has proved that the biosorbent surface was damaged after being in contact with copper (waste), implying that heavy metals (molecules) attach to different functional cell surfaces and change the biosorbent surface. The adsorption process of this research follows Langmuir Isotherm with the R2 value close to 1. The immobilized mixed culture biosorbent is capable of optimally removing copper at temperature of 30°C and initial Cu2+ concentration of 17.62 mg/l.

  20. Barium borosilicate glass - a potential matrix for immobilization of sulfate bearing high-level radioactive liquid waste

    International Nuclear Information System (INIS)

    Kaushik, C.P.; Mishra, R.K.; Sengupta, P.; Kumar, Amar; Das, D.; Kale, G.B.; Raj, Kanwar

    2006-01-01

    Borosilicate glass formulations adopted worldwide for immobilization of high-level radioactive liquid waste (HLW) is not suitable for sulphate bearing HLW, because of its low solubility in such glass. A suitable glass matrix based on barium borosilicate has been developed for immobilization of sulphate bearing HLW. Various compositions based on different glass formulations were made to examine compatibility with waste oxide with around 10 wt% sulfate content. The vitrified waste product obtained from barium borosilicate glass matrix was extensively evaluated for its characteristic properties like homogeneity, chemical durability, glass transition temperature, thermal conductivity, impact strength, etc. using appropriate techniques. Process parameters like melt viscosity and pour temperature were also determined. It is found that SB-44 glass composition (SiO 2 : 30.5 wt%, B 2 O 3 : 20.0 wt%, Na 2 O: 9.5 wt% and BaO: 19.0 wt%) can be safely loaded with 21 wt% waste oxide without any phase separation. The other product qualities of SB-44 waste glass are also found to be on a par with internationally adopted waste glass matrices. This formulation has been successfully implemented in plant scale

  1. Setting and Stiffening of Cementitious Components in Cast Stone Waste Form for Disposal of Secondary Wastes from the Hanford waste treatment and immobilization plant

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chul-Woo; Chun, Jaehun; Um, Wooyong; Sundaram, S. K.; Westsik, Joseph H.

    2013-04-01

    Cast stone is a cementitious waste form, a viable option to immobilize secondary nuclear liquid wastes generated from Hanford vitrification plant. While the strength and radioactive technetium leaching of different waste form candidates have been reported, no study has been performed to understand the flow and stiffening behavior of Cast Stone, which is essential to ensure the proper workability, especially considering necessary safety as a nuclear waste form in a field scale application. The rheological and ultrasonic wave reflection (UWR) measurements were used to understand the setting and stiffening Cast Stone batches. X-ray diffraction (XRD) was used to find the correlation between specific phase formation and the stiffening of the paste. Our results showed good correlation between rheological properties of the fresh Cast Stone mixture and phase formation during hydration of Cast Stone. Secondary gypsum formation originating from blast furnace slag was observed in Cast Stone made with low concentration simulants. The formation of gypsum was suppressed in high concentration simulants. It was found that the stiffening of Cast Stone was strongly dependent on the concentration of simulant. A threshold concentration for the drastic change in stiffening was found at 1.56 M Na concentration.

  2. Recharge Data Package for the Immobilized Low-Activity Waste 2001 Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    MJ Fayer; EM Murphy; JL Downs; FO Khan; CW Lindenmeier; BN Bjornstad

    2000-01-18

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are currently stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as immobilized low-activity waste (ILAW) is to vitrify the waste and place the product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is known as the Hanford ILAW Performance Assessment (PA) Activity, hereafter called the ILAW PA project. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require predictions of contaminant migration from the facility. To make such predictions will require estimates of the fluxes of water moving through the sediments within the vadose zone around and beneath the disposal facility. These fluxes, loosely called recharge rates, are the primary mechanism for transporting contaminants to the groundwater. Pacific Northwest National Laboratory (PNNL) assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of recharge rates for current conditions and long-term scenarios involving the shallow-land disposal of ILAW. Specifically, recharge estimates are needed for a filly functional surface cover; the cover sideslope, and the immediately surrounding terrain. In addition, recharge estimates are needed for degraded cover conditions. The temporal scope of the analysis is 10,000 years, but could be longer if some contaminant peaks occur after 10,000 years. The elements of this report compose the Recharge Data Package, which provides estimates of recharge rates for the scenarios being considered in the 2001 PA. Table S.1 identifies the surface features and

  3. Recharge Data Package for the Immobilized Low-Activity Waste 2001 Performance Assessment

    International Nuclear Information System (INIS)

    MJ Fayer; EM Murphy; JL Downs; FO Khan; CW Lindenmeier; BN Bjornstad

    2000-01-01

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are currently stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as immobilized low-activity waste (ILAW) is to vitrify the waste and place the product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is known as the Hanford ILAW Performance Assessment (PA) Activity, hereafter called the ILAW PA project. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require predictions of contaminant migration from the facility. To make such predictions will require estimates of the fluxes of water moving through the sediments within the vadose zone around and beneath the disposal facility. These fluxes, loosely called recharge rates, are the primary mechanism for transporting contaminants to the groundwater. Pacific Northwest National Laboratory (PNNL) assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of recharge rates for current conditions and long-term scenarios involving the shallow-land disposal of ILAW. Specifically, recharge estimates are needed for a filly functional surface cover; the cover sideslope, and the immediately surrounding terrain. In addition, recharge estimates are needed for degraded cover conditions. The temporal scope of the analysis is 10,000 years, but could be longer if some contaminant peaks occur after 10,000 years. The elements of this report compose the Recharge Data Package, which provides estimates of recharge rates for the scenarios being considered in the 2001 PA. Table S.1 identifies the surface features and

  4. Immobilization of simulated low and intermediate level waste in alkali-activated slag-fly ash-metakaolin hydroceramics

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jin, E-mail: wjin761026@163.com [State Key Laboratory Cultivation Base for Nonmetal Composite and Functional Materials, Southwest University of Science and Technology, Mianyang 621010, Sichuan (China); School of Materials Science and Engineering, Southwest University of Science and Technology, Mianyang 621010, Sichuan (China); Wang, Jun-xia; Zhang, Qin [School of Materials Science and Engineering, Southwest University of Science and Technology, Mianyang 621010, Sichuan (China); Li, Yu-xiang [State Key Laboratory Cultivation Base for Nonmetal Composite and Functional Materials, Southwest University of Science and Technology, Mianyang 621010, Sichuan (China); School of Materials Science and Engineering, Southwest University of Science and Technology, Mianyang 621010, Sichuan (China)

    2016-04-15

    Highlights: • Evaluation of the suitability of ASFMH for solidifying simulated S-LILW. • The introduction of S-LILW avails forming zeolitic phases of ASFMH waste forms. • The ASFMH waste forms have low leachability and high compressive strength. - Abstract: In the current study, the alkali-activated slag-fly ash-metakaolin hydroceramic (ASFMH) waste forms for immobilizing simulated low and intermediate level waste (S-LILW) were prepared by hydrothermal process. The crystalline phase compositions, morphology, compressive strength and aqueous stability of S-LILW ASFMH waste forms were investigated. The results showed that the main crystalline phases of S-LILW ASFMH waste forms were analcime and zeolite NaP1. The changes of Si/Al molar ratio (from 1.7 to 2.2) and Ca/Al molar ratio (from 0.15 to 0.35) had little effect on the phase compositions of S-LILW ASFMH waste forms. However, the hydrothermal temperature, time as well as the content of S-LILW (from 12.5 to 37.5 wt%) had a major impact on the phase compositions. The compressive strength of S-LILW ASFMH waste forms was not less than 20 MPa when the content of S-LILW reached 37.5 wt%. In addition, the aqueous stability testing was carried out using the standard MCC-1 static leach test method; the normalized elemental leach rates of Sr and Cs were fairly constant in a low value below 5 × 10{sup −4} g m{sup −2} d{sup −1} and 3 × 10{sup −4} g m{sup −2} d{sup −1} after 28 days, respectively. It is indicated that ASFMH waste form could be a potential host for safely immobilizing LILW.

  5. Production and characterization of red mud based on glasses for the immobilization of nuclear wastes

    International Nuclear Information System (INIS)

    Vieira, Heveline

    2015-01-01

    Glasses based on red mud, a residual material from bauxite processing, were developed and characterized in this work. In order to promote its use, a minimum 60 wt% of red mud was used in the production of the glasses. According to XRD results, materials containing considerable amorphous phases were produced when using red mud as raw material. These amorphous phases were observed even though crystalline phases associated to Fe coming from the red mud itself were present. The material denominated 60L40S, which has a nominal composition of 60 wt% red mud showed the best properties comparing with the others compositions studied. However, these materials presented a high melting temperature. Changes in the composition of this material were made with the objective of lowering this temperature. Results indicated that the changes made to the material were successful in the reduction of the melting temperature. However, a reduction in the chemical properties of the resulting material was observed. Elements usually found in the chemical composition of nuclear wastes were added to the glasses produced. It was done with the objective of determining the effect of these elements on the chemical and physical properties of the red mud based glasses obtained. It was found that it was possible to add up to 15 wt% of these elements to the materials produced. The addition of these simulant materials promoted a reduction in the melting temperature of the resulting material. Above 15 wt%, the added elements precipitate in the structure of the resulting material. Even though the reduction in the chemical durability of the 60L40S material when simulant elements were added, it was observed that this material contained the simulant elements confined in its structure when in contact with water. This is a promising result, since it indicates that the 60L40S has the potential to immobilize elements from nuclear wastes . (author)

  6. Waste treatment and immobilization technologies involving inorganic sorbents. Final report of a co-ordinated research programme 1992-1996

    International Nuclear Information System (INIS)

    1997-06-01

    A Coordinated Research Programme (CRP) for the application of inorganic sorbents in liquid waste treatment and immobilization was initiated by the IAEA in 1992. The results of this CRP are presented in this report. Fifteen institutions from fourteen countries were involved in this programme. The framework of this CRP was: (1) to conduct fundamental studies on sorbent structure and sorption mechanism; (2) to obtain thermodynamic and kinetic data of the treatment process; (3) to define sorption mechanism of radionuclides on different soils; (4) to identify sorbents appropriate for treatment of liquid waste streams; (5) to develop standard tests to be able to compare results of different groups of investigations. Refs, figs, tabs

  7. Study on rich alumina alkali-activated slag clay minerals cementitious materials for immobilization of radioactive waste

    International Nuclear Information System (INIS)

    Li Yuxiang; Qian Guangren; Yi Facheng; Shi Rongming; Fu Yibei; Li Lihua; Zhang Jun

    1999-01-01

    The composition and some properties of its pastes of rich alumina alkali-activated slag clay minerals (RAAASCM) cementitious materials for immobilization of radioactive waste are studied. Experimental results show that heat activated kaolinite, Xingjiang zeolite, modified attapulgite clay are better constituents of RAAASCM. RAAASCM cementitious materials pastes exhibit high strength, low porosity, fewer harmful pore, and high resistance to sulphate corrosion as well as gamma irradiation. The Sr 2+ , Cs + leaching portion of the simulated radioactive waste forms based on RAAASCM, is low

  8. Gamma radiolysis effects on leaching behavior of ceramic materials for nuclear fuel waste immobilization containers

    International Nuclear Information System (INIS)

    Onofrei, M.; Raine, D.K.; Hocking, W.H.; George, K.; Betteridge, J.S.

    1986-01-01

    The leaching behavior of ceramic materials for nuclear fuel waste immobilization containers, under the influence of a moderate gamma dose rate (4 Gy/h), has been investigated. Samples of Al/sub 2/O/sub 3/, stabilized ZrO/sub 2/, TiO/sub 2/, cermet (70% Al/sub 2/O-30% TiC), porcelain (with high Al/sub 2/O/sub 3/ content), and concrete (with sulfate-resisting portland cement plus silica fume) have been leached in Standard Canadian Shield Saline Solution (SCSSS), and SCSSS plus clay and sand (components of the disposal system), at 100 0 and 150 0 C for 231 and 987 days, respectively. Leaching solutions were analyzed and the surfaces of the leached samples were investigated by scanning electron microscopy in conjunction with energy dispersive X-ray spectroscopy and secondary ion mass spectrometry. Radiolysis did not appear to enhance the leaching, with or without bentonite and sand in the system. Analysis of the gas phase from sealed capsules showed O/sub 2/ depletion and production of CO/sub 2/ in all experiments containing bentonite. The decrease in O/sub 2/ is attributed to the leaching from the clay of Fe(II) species, which can participate in redox reactions with radicals generated by radiolysis. The CO/sub 2/ is produced from either the organic or inorganic fraction in the bentonite

  9. Systematic investigation of the strontium zirconium phosphate ceramic form for nuclear waste immobilization

    Science.gov (United States)

    Pet'kov, Vladimir; Asabina, Elena; Loshkarev, Vladimir; Sukhanov, Maksim

    2016-04-01

    We have summarized our data and literature ones on the thermophysical properties and hydrolytic stability of Sr0.5Zr2(PO4)3 compound as a host NaZr2(PO4)3-type (NZP) structure for immobilization of 90Sr-containing radioactive waste. Absence of any polymorphic transformations on the temperature dependence of its heat capacity between 7 and 665 K is caused by the stability of crystalline Sr0.5Zr2(PO4)3. Calculated values of thermal conductivity coefficients at zero porosity in the range 298-673 K were 1.86-2.40 W·m-1 K-1. The compound may be classified as low thermal expanding material due to its average linear thermal expansion coefficient. Study of the hydrolytic stability in acid and alkaline media has shown that the relative mass fraction of Sr2+ ions, released into aggressive leaching media, didn't exceed 1% of the mass of sample. Soxhlet leaching studies have shown substantial resistance towards the release of Sr2+ ions into distilled water. Feeble sinterability constrains practical applications of NZP substances, that is why known in literature methods of Sr0.5Zr2(PO4)3 dense ceramics obtaining have been reviewed.

  10. The integrated criticality safety evaluation for the Hanford tank waste treatment and immobilization plant

    International Nuclear Information System (INIS)

    Losey, D. C.; Miles, R. E.; Perks, M. F.

    2009-01-01

    The Criticality Safety Evaluation Report (CSER) for the Hanford Tank Waste Treatment and Immobilization Plant (WTP) has been developed as a single, integrated evaluation with a scope that covers all of the planned WTP operations. This integrated approach is atypical, as the scopes of criticality evaluations are usually more narrowly defined. Several adjustments were made in developing the WTP CSER, but the primary changes were to provide introductory overview for the criticality safety control strategy and to provide in-depth analysis of the underlying physical and chemical mechanisms that contribute to ensuring safety. The integrated approach for the CSER allowed a more consistent evaluation of safety and avoided redundancies that occur when evaluation is distributed over multiple documents. While the approach used with the WTP CSER necessitated more coordination and teamwork, it has yielded a report is that more integrated and concise than is typical. The integrated approach with the CSER produced a simple criticality control scheme that uses relatively few controls. (authors)

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

    International Nuclear Information System (INIS)

    Vieira, Vanessa Mota; de Tello, Cledola Cassia Oliveira

    2013-01-01

    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)

  12. Hanford Immobilized Low Activity Waste (ILAW) Performance Assessment 2001 Version [Formerly DOE/RL-97-69] [SEC 1 & 2

    Energy Technology Data Exchange (ETDEWEB)

    MANN, F.M.

    2000-08-01

    The Hanford Immobilized Low-Activity Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-activity fraction of waste presently contained in Hanford Site tanks. The tank waste is the byproduct of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste is stored in underground single- and double-shell tanks. The tank waste is to be retrieved, separated into low-activity and high-level fractions, and then immobilized by vitrification. The US. Department of Energy (DOE) plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at the Hanford Site until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to modify the current Disposal Authorization Statement for the Hanford Site that would allow the following: construction of disposal trenches; and filling of these trenches with ILAW containers and filler material with the intent to dispose of the containers.

  13. Silicate Based Glass Formulations for Immobilization of U.S. Defense Wastes Using Cold Crucible Induction Melters

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L.; Kim, Dong-Sang; Schweiger, Michael J.; Marra, James C.; Lang, Jesse B.; Crum, Jarrod V.; Crawford, Charles L.; Vienna, John D.

    2014-05-22

    The cold crucible induction melter (CCIM) is an alternative technology to the currently deployed liquid-fed, ceramic-lined, Joule-heated melter for immobilizing of U.S. tank waste generated from defense related reprocessing. In order to accurately evaluate the potential benefits of deploying a CCIM, glasses must be developed specifically for that melting technology. Related glass formulation efforts have been conducted since the 1990s including a recent study that is first documented in this report. The purpose of this report is to summarize the silicate base glass formulation efforts for CCIM testing of U.S. tank wastes. Summaries of phosphate based glass formulation and phosphate and silicate based CCIM demonstration tests are reported separately (Day and Ray 2013 and Marra 2013, respectively). Combined these three reports summarize the current state of knowledge related to waste form development and process testing of CCIM technology for U.S. tank wastes.

  14. Cement mortar-degraded spinney waste composite as a matrix for immobilizing some low and intermediate level radioactive wastes: Consistency under frost attack

    International Nuclear Information System (INIS)

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

    2012-01-01

    Highlights: ► Spinney fiber is one of the wastes generated from spinning of cotton raw materials. ► Cement mortar composite was hydrated by using the degraded slurry of spinney wastes. ► Frost resistance was assessed for the mortar-degraded spinney waste composite specimens. ► SEM image, FT-IR and XRD patterns were performed for samples subjected to frost attack. - Abstract: The increasing amounts of spinning waste fibers generated from cotton fabrication are problematic subject. Simultaneous shortage in the landfill disposal space is also the most problem associated with dumping of these wastes. Cement mortar composite was developed by hydrating mortar components using the waste slurry obtained from wet oxidative degradation of these spinney wastes. The consistency of obtained composite was determined under freeze–thaw events. Frost resistance was assessed for the mortar composite specimens by evaluating its compressive strength, apparent porosity and mass loss at the end of each period of freeze–thaw up to 45 cycles. Scanning electron microscopy, infrared spectroscopy and X-ray diffraction analyses were performed for samples subjected to frost attack aiming at evaluating the cement mortar in the presence of degraded spinney waste. The cement mortar composite exhibits acceptable resistance and durability against the freeze–thaw treatment that could be chosen in radioactive waste management as immobilizing agent for some low and intermediate level radioactive wastes.

  15. Preliminary Technology Maturation Plan for Immobilization of High-Level Waste in Glass Ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, John D.; Crum, Jarrod V.; Sevigny, Gary J.; Smith, G L.

    2012-09-30

    A technology maturation plan (TMP) was developed for immobilization of high-level waste (HLW) raffinate in a glass ceramics waste form using a cold-crucible induction melter (CCIM). The TMP was prepared by the following process: 1) define the reference process and boundaries of the technology being matured, 2) evaluate the technology elements and identify the critical technology elements (CTE), 3) identify the technology readiness level (TRL) of each of the CTE’s using the DOE G 413.3-4, 4) describe the development and demonstration activities required to advance the TRLs to 4 and 6 in order, and 5) prepare a preliminary plan to conduct the development and demonstration. Results of the technology readiness assessment identified five CTE’s and found relatively low TRL’s for each of them: • Mixing, sampling, and analysis of waste slurry and melter feed: TRL-1 • Feeding, melting, and pouring: TRL-1 • Glass ceramic formulation: TRL-1 • Canister cooling and crystallization: TRL-1 • Canister decontamination: TRL-4 Although the TRL’s are low for most of these CTE’s (TRL-1), the effort required to advance them to higher values. The activities required to advance the TRL’s are listed below: • Complete this TMP • Perform a preliminary engineering study • Characterize, estimate, and simulate waste to be treated • Laboratory scale glass ceramic testing • Melter and off-gas testing with simulants • Test the mixing, sampling, and analyses • Canister testing • Decontamination system testing • Issue a requirements document • Issue a risk management document • Complete preliminary design • Integrated pilot testing • Issue a waste compliance plan A preliminary schedule and budget were developed to complete these activities as summarized in the following table (assuming 2012 dollars). TRL Budget Year MSA FMP GCF CCC CD Overall $M 2012 1 1 1 1 4 1 0.3 2013 2 2 1 1 4 1 1.3 2014 2 3 1 1 4 1 1.8 2015 2 3 2 2 4 2 2.6 2016 2 3 2 2 4 2 4

  16. DEPARTMENT OF ENERGY (DOE) MANAGEMENT OF THE HANFORD WASTE TREATMENT and IMMOBILIZATION PLANT

    International Nuclear Information System (INIS)

    SHRADER, T.A.

    2005-01-01

    The US Department of Energy Office of River Protection is currently overseeing the construction of the new Hanford Site Waste Treatment and Immobilization Plant (more commonly referred to as the Waste Treatment Plant). In December 2000, a contract was awarded to Bechtel National, Inc. for the design, construction, and commissioning of the $5.8 billion facility to treat and vitrify a significant portion of the waste currently stored in large underground tanks on the Hanford Site. As the owner, the Office of River Protection has developed an organization to oversee the design, construction, and commissioning of the facility. A Federal Project Director is responsible for all aspects of the project, including safety, design, construction, commissioning; and the baseline (scope, cost, and schedule). The Project Director reports to the Manager of the Office of River Protection and recommends changes to the contract requirements, safety basis documents, or the baseline. Approximately 30 engineers, scientists, and other support personnel have been assigned to a unique organization that supports the Federal Project Director in providing oversight of each phase of the project (i.e., design, construction, and commissioning). The organization includes an Engineering Division, a Programs and Projects Division, a Safety Authorization Basis Team, and an Operations and Commissioning Team. This organization is unique within the Department of Energy and provides a focused team to resolve issues of safety, cost, schedule, technical design changes, and construction. This paper will describe this team and show how the Office of River Protection utilizes this oversight team to manage this complex, accelerated project. The size and technical complexity of the facility poses unique challenges for safety, permitting, commissioning, engineering, and baseline control. A robust training and qualification program has been developed that will insure the Departmental personnel working closely

  17. Stabilization of organic matter and nitrogen immobilization during mechanical-biological treatment and landfilling of residual municipal solid waste

    International Nuclear Information System (INIS)

    Heiss-Ziegler, C.

    2000-04-01

    Synthesis of humic substances and nitrogen immobilization during mechanical-biological treatment of waste and the behavior of biologically stabilized waste under anaerobic landfill conditions were investigated. Samples were taken from a large-scale treatment plant. Anaerobic conditions were simulated in lab scale test cells. Humic substances were analyzed photometrically and gravimetrically. The nitrogen immobilization was investigated by sequential leaching tests and by analyzing the non acid hydrolyzable nitrogen. Humic acids were mainly synthesized during the beginning of the intensive rotting phase. Later on in the process no significant changes occurred. The humic acid content rose up to 6,8 % DS org. It correlated well with the stability parameters respiration activity and accumulated gas production. In the coarse of the treatment the nitrogen load emitted during the consecutive leaching tests dropped from 50 % down to less than 20 % total nitrogen. The non acid hydrolyzable nitrogen rose from 17 up to 42 % Kjeldahl nitrogen content. Nevertheless the mechanical-biological treatment is not significantly shortening the aftercare period of a landfill concerning liquid nitrogen emissions. The reduced nitrogen emission potential is released more slowly. When reactive waste material was exposed to anaerobic conditions, humic and fulvic acids were synthesized up to the point when intensive gas production started and then were remineralized. Stabilized waste materials after treatment of various intensity behaved differently under anaerobic conditions. Steady and decreasing humic acid contents were observed. (author)

  18. Calcium-borosilicate glass-ceramics wasteforms to immobilize rare-earth oxide wastes from pyro-processing

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Miae [Department of Materials Science and Engineering and Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), Pohang, Gyeongbuk, 790-784 (Korea, Republic of); Heo, Jong, E-mail: jheo@postech.ac.kr [Department of Materials Science and Engineering and Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), Pohang, Gyeongbuk, 790-784 (Korea, Republic of); Department of Materials Engineering, Adama Science and Technology University (ASTU), PO Box 1888, Adama (Ethiopia)

    2015-12-15

    Glass-ceramics containing calcium neodymium(cerium) oxide silicate [Ca{sub 2}Nd{sub 8-x}Ce{sub x}(SiO{sub 4}){sub 6}O{sub 2}] crystals were fabricated for the immobilization of radioactive wastes that contain large portions of rare-earth ions. Controlled crystallization of alkali borosilicate glasses by heating at T ≥ 750 °C for 3 h formed hexagonal Ca–silicate crystals. Maximum lanthanide oxide waste loading was >26.8 wt.%. Ce and Nd ions were highly partitioned inside Ca–silicate crystals compared to the glass matrix; the rare-earth wastes are efficiently immobilized inside the crystalline phases. The concentrations of Ce and Nd ions released in a material characterization center-type 1 test were below the detection limit (0.1 ppb) of inductively coupled plasma mass spectroscopy. Normalized release values performed by a product consistency test were 2.64·10{sup −6} g m{sup −2} for Ce ion and 2.19·10{sup −6} g m{sup −2} for Nd ion. Results suggest that glass-ceramics containing calcium neodymium(cerium) silicate crystals are good candidate wasteforms for immobilization of lanthanide wastes generated by pyro-processing. - Highlights: • Glass-ceramic wasteforms containing Ca{sub 2}Nd{sub 8-x}Ce{sub x}(SiO{sub 4}){sub 6}O{sub 2} crystals were synthesized to immobilize lanthanide wastes. • Maximum lanthanide oxide waste loading was >26.8 wt.%. • Ce and Nd ions were highly partitioned inside Ca–Nd–silicate crystals compared to glass matrix. • Amounts of Ce and Nd ions released in the material characterization center-type 1 were below the detection limit (0.1 ppb). • Normalized release values performed by a PCT were 2.64• 10{sup −6} g m{sup −2} for Ce ions and 2.19• 10{sup −6} g m{sup −2} for Nd ions.

  19. Development of an immobilization process for heavy metal containing galvanic solid wastes by use of sodium silicate and sodium tetraborate

    Energy Technology Data Exchange (ETDEWEB)

    Aydın, Ahmet Alper, E-mail: ahmetalperaydin@gmail.com [Chair of Urban Water Systems Engineering, Technische Universität München, Am Coulombwall, 85748 Garching (Germany); Aydın, Adnan [Istanbul Bilim University, School of Health, Esentepe, Istanbul, Sisli, 34394 (Turkey)

    2014-04-01

    Highlights: • A new physico-chemical process below 1000 °C for immobilization of galvanic sludges. • Sodium tetraborate and sodium silicate have been used as additives. • A strategy for adjustment of solid waste/additive mixture composition is presented. • Strategy is valid for wastes of hydrometallurgical and electro-plating processes. • Lower energy consumption and treated waste volume, shorter process time are provided. - Abstract: Heavy metal containing sludges from wastewater treatment plants of electroplating industries are designated as hazardous waste since their improper disposal pose high risks to environment. In this research, heavy metal containing sludges of electroplating industries in an organized industrial zone of Istanbul/Turkey were used as real-sample model for development of an immobilization process with sodium tetraborate and sodium silicate as additives. The washed sludges have been precalcined in a rotary furnace at 900 °C and fritted at three different temperatures of 850 °C, 900 °C and 950 °C. The amounts of additives were adjusted to provide different acidic and basic oxide ratios in the precalcined sludge-additive mixtures. Leaching tests were conducted according to the toxicity characteristic leaching procedure Method 1311 of US-EPA. X-ray diffraction (XRD), X-ray fluorescence (XRF), scanning electron microscope-energy dispersive spectrometer (SEM-EDS) and flame atomic absorption spectroscopy (FAAS) have been used to determine the physical and chemical changes in the products. Calculated oxide molar ratios in the precalcined sludge-additive mixtures and their leaching results have been used to optimize the stabilization process and to determine the intervals of the required oxide ratios which provide end-products resistant to leaching procedure of US-EPA. The developed immobilization-process provides lower energy consumption than sintering-vitrification processes of glass–ceramics.

  20. The thorium phosphate diphosphate as matrix for radioactive waste conditioning: radionuclide immobilization and behavior under irradiation

    International Nuclear Information System (INIS)

    Pichot, Erwan

    1999-01-01

    The aim of this work was to perform successively the decontamination of liquid solutions and the final immobilization of radionuclide storage using the same matrix. For this, thorium phosphate-diphosphate (TPD) of the formula Th 4 P 6 O 23 , is proposed as a very resistant to water corrosion matrix. A new compound, thorium phosphate hydrogeno-phosphate (TPHP) of the formula Th 2 (PO 4 ) 2 (HPO 4 ), nH 2 O with n=3-7 was synthesized and characterized. Heated at 1100 deg.C it is transformed into the TDP. Ion exchange properties of TPHP were investigated. The exchange yields of imponderable caesium, strontium and americium ion onto TPHP (NaNO 3 0.1 M media at pH=6) are equal to 60% for the first one and 100% for the two others. The results interpreted in terms of ion-exchange led to determine selectivity coefficient values for each cation and suggested that only hydrated ions are exchanged. While the TPD is proposed for the high level nuclear waste storage, the irradiation effects, particularly structural modifications were studied using both γ irradiation and charged particle irradiation. ESR and TL methods were carried out in order to identify radicals created during gamma radiation exposure. Correlation between ESR and TL experiments performed at room temperature clearly show three of PO 3 2- species and one POO· species of free radicals. We have shown that Au-ion irradiation in the range of MeV energy involved TPD structure and chemical modifications. Important sputtering was interpreted in terms of local thermal chemical decomposition. We have shown, at room temperature, that the amorphization dose for heavy ion irradiation is between 0.1 to 0.4 dpa. (author)

  1. APPROVAL OF WASTE TREATMENT AND IMMOBILIZATION PLANT CONTRACTOR-INITIATED AUTHORIZATION BASIS AMENDMENT REQUESTS (ABAR)

    International Nuclear Information System (INIS)

    JONES GL

    2008-01-01

    The objective is to describe the process used by the Office of River Protection (ORP) for evaluating and implementing Contractor-initiated changes to the Waste Treatment and Immobilization Plant (WTP) Authorization Basis (AB). The WTP Project's history has provided a unique challenge for establishing and maintaining an ORP-approved AB during design and construction. Until operations begin, the project cannot implement the classic Unreviewed Safety Question (USQ) process to determine when ORP approval of Contractor-initiated changes is required. A 'quasiUSQ' process has been implemented that defines when AB changes could occur. The three types of AB changes are (1) Limited Scope Changes, (2) Authorization Basis Deviations, and (3) Authorization Basis Amendment Request (ABAR). DOE RL/REG 97-13, 'Office of River Protection Position on Contractor-Initiated Changes to the Authorization Basis', describes the process the WTP Contractor must follow to make changes to the AB, with and without ORP approval. The process uses a 'safety evaluation' process that is similar to the USQ process but at a more qualitative level. The maturation of the WTP Contractor's facility design and activities, and other changing conditions, resulted in a process that allows the Contractor to make changes to the AB without ORP approval; however, those changes that may significantly affect nuclear safety do require ORP approval. This process balances the WTP regulatory principle of efficiency with assurance that adequate safety will not be compromised. The process has reduced the number of ABARs requiring ORP approval and reduced the potential for delays in design and procurement activities

  2. Glass-bonded iodosodalite waste form for immobilization of 129 I

    Energy Technology Data Exchange (ETDEWEB)

    Chong, Saehwa; Peterson, Jacob A.; Riley, Brian J.; Tabada, Diana; Wall, Donald; Corkhill, Claire L.; McCloy, John S.

    2018-06-01

    Immobilization of radioiodine (e.g., 129I, 131I) is an important need for current and future nuclear fuel cycles. For the current work, iodosodalite [Na8(AlSiO4)6I2] was synthesized hydrothermally from metakaolin, NaI, and NaOH. Following hydrothermal treatment, dried unwashed powders were used to make glass-bonded iodosodalite waste forms by heating pressed pellets at 650, 750, or 850 °C with two different types of sodium borosilicate glass binders, i.e., NBS-4 and SA-800. These heat-treated specimens were characterized with X-ray diffraction, Fourier-transform infrared spectroscopy, scanning electron microscopy, energy dispersive spectroscopy, thermal analysis, porosity and density measurements, neutron activation analysis, and inductively-coupled plasma mass spectrometry. The pellets mixed with 10 mass% of NBS-4 or SA-800 and heat-treated at 750 °C contained relatively high percentage iodine retention (~44-47 % of the maximum iodine loading) with relatively low porosities, while other pellets with higher percentages iodine retention either contained higher porosity or were not completely sintered. ASTM C1308 chemical durability tests of monolithic specimens showed a large initial release of Na, Al, Si, and I on the first day, possibly from water-soluble salt crystals or non-durable amorphous phases. Release rates of Na and Si were higher than for Al and I, probably due to a poorly durable Na-Si-O phase from the glass bonding matrix. The cumulative normalized release of iodine was 12.5 g m-2 for the first 10 1-d exchanges, suggestive of coherent dissolution. The average release rate from 10-24 days during the 7-d exchange intervals was 0.2336 g m-2 d-1.

  3. Physical modeling of joule heated ceramic glass melters for high level waste immobilization

    International Nuclear Information System (INIS)

    Quigley, M.S.; Kreid, D.K.

    1979-03-01

    This study developed physical modeling techniques and apparatus suitable for experimental analysis of joule heated ceramic glass melters designed for immobilizing high level waste. The physical modeling experiments can give qualitative insight into the design and operation of prototype furnaces and, if properly verified with prototype data, the physical models could be used for quantitative analysis of specific furnaces. Based on evaluation of the results of this study, it is recommended that the following actions and investigations be undertaken: It was not shown that the isothermal boundary conditions imposed by this study established prototypic heat losses through the boundaries of the model. Prototype wall temperatures and heat fluxes should be measured to provide better verification of the accuracy of the physical model. The VECTRA computer code is a two-dimensional analytical model. Physical model runs which are isothermal in the Y direction should be made to provide two-dimensional data for more direct comparison to the VECTRA predictions. The ability of the physical model to accurately predict prototype operating conditions should be proven before the model can become a reliable design tool. This will require significantly more prototype operating and glass property data than were available at the time of this study. A complete set of measurements covering power input, heat balances, wall temperatures, glass temperatures, and glass properties should be attempted for at least one prototype run. The information could be used to verify both physical and analytical models. Particle settling and/or sludge buildup should be studied directly by observing the accumulation of the appropriate size and density particles during feeding in the physical model. New designs should be formulated and modeled to minimize the potential problems with melter operation identifed by this study

  4. Evaluation of support matrices for immobilization of anaerobic consortia for efficient carbon cycling in waste regeneration.

    Science.gov (United States)

    Chauhan, Ashvini; Ogram, Andrew

    2005-02-18

    Efficient metabolism of fatty acids during anaerobic waste digestion requires development of consortia that include "fatty acid consuming H(2) producing bacteria" and methanogenic bacteria. The objective of this research was to optimize methanogenesis from fatty acids by evaluating a variety of support matrices for use in maintaining efficient syntrophic-methanogenic consortia. Tested matrices included clays (montmorillonite and bentonite), glass beads (106 and 425-600mum), microcarriers (cytopore, cytodex, cytoline, and cultispher; conventionally employed for cultivation of mammalian cell lines), BioSep beads (powdered activated carbon), and membranes (hydrophilic; nylon, polysulfone, and hydrophobic; teflon, polypropylene). Data obtained from headspace methane (CH(4)) analyses as an indicator of anaerobic carbon cycling efficiency indicated that material surface properties were important in maintenance and functioning of the anaerobic consortia. Cytoline yielded significantly higher CH(4) than other matrices as early as in the first week of incubation. 16S rRNA gene sequence analysis from crushed cytoline matrix showed the presence of Syntrophomonas spp. (butyrate oxidizing syntrophs) and Syntrophobacter spp. (propionate oxidizing syntrophs), with Methanosaeta spp. (acetate utilizing methanogen), and Methanospirillum spp. (hydrogen utilizing methanogen) cells. It is likely that the more hydrophobic surfaces provided a suitable surface for adherence of cells of syntrophic-methanogenic consortia. Cytoline also appeared to protect entrapped consortia from air, resulting in rapid methanogenesis after aerial exposure. Our study suggests that support matrices can be used in anaerobic digestors, pre-seeded with immobilized or entrapped consortia on support matrices, and may be of value as inoculant-adsorbents to rapidly initiate or recover proper system functioning following perturbation.

  5. Inhibition of acid mine drainage and immobilization of heavy metals from copper flotation tailings using a marble cutting waste

    Science.gov (United States)

    Tozsin, Gulsen

    2016-01-01

    Acid mine drainage (AMD) with high concentrations of sulfates and metals is generated by the oxidation of sulfide bearing wastes. CaCO3-rich marble cutting waste is a residual material produced by the cutting and polishing of marble stone. In this study, the feasibility of using the marble cutting waste as an acid-neutralizing agent to inhibit AMD and immobilize heavy metals from copper flotation tailings (sulfide- bearing wastes) was investigated. Continuous-stirring shake-flask tests were conducted for 40 d, and the pH value, sulfate content, and dissolved metal content of the leachate were analyzed every 10 d to determine the effectiveness of the marble cutting waste as an acid neutralizer. For comparison, CaCO3 was also used as a neutralizing agent. The average pH value of the leachate was 2.1 at the beginning of the experiment ( t = 0). In the experiment employing the marble cutting waste, the pH value of the leachate changed from 6.5 to 7.8, and the sulfate and iron concentrations decreased from 4558 to 838 mg/L and from 536 to 0.01 mg/L, respectively, after 40 d. The marble cutting waste also removed more than 80wt% of heavy metals (Cd, Cr, Cu, Ni, Pb, and Zn) from AMD generated by copper flotation tailings.

  6. Application of self-propagation high-temperature synthesis for immobilization of hard radioactive wastes in ceramet materials

    International Nuclear Information System (INIS)

    Ilyin, E.; Pashkeev, I.; Senin, A.; Gerasimova, N.

    2001-01-01

    The possibility of self-propagating high-temperature synthesis (SPHTS) application for an immobilization of solid high level wastes (HLW) in cermet materials is considered. The schemes of multilayer cermet blocks formation are offered. Such blocks consist of a ceramet core with immobilized HLW and a protective cover - ceramet without HLW. The influence of the base components form (pure Ti and Si, ferrotitanium and ferrosilicon), metallic components (Ni, Cu, Cr, Fe, ferrochromium) and nonmetallic components (SiO 2 , Al 2 O 3 , TiO 2 ) on burning rate and cover ceramet structure is investigated in compositions on a basis of Ti+B, Ti+Si, Ti+C systems. Model samples of multilayer cermet blocks are manufactured using of HLW simulators. (authors)

  7. Immobilization of tritiated waste-water by hydraulic cements. A survey of the state-of-the-art

    International Nuclear Information System (INIS)

    Mannone, F.

    1987-01-01

    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

  8. Durability, mechanical, and thermal properties of experimental glass-ceramic forms for immobilizing ICPP high level waste

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1990-01-01

    The high-level liquid waste generated at the Idaho Chemical Processing Plant (ICPP) is routinely solidified into granular calcined high-level waste (HLW) and stored onsite. Research is being conducted at the ICPP on methods of immobilizing the HLW, including developing a durable glass-ceramic form which has the potential to significantly reduce the final waste volume by up to 60% compared to a glass form. Simulated, pilot plant, non-radioactive, calcines similar to the composition of the calcined HLW and glass forming additives are used to produce experimental glass-ceramic forms. The objective of the research reported in this paper is to study the impact of ground calcine particle size on durability and mechanical and thermal properties of experimental glass-ceramic forms

  9. Drug product immobilization in recycled polyethylene/polypropylene reclaimed from municipal solid waste: experimental and numerical assessment.

    Science.gov (United States)

    Saad, Walid; Slika, Wael; Mawla, Zara; Saad, George

    2017-12-01

    Recently, there has been a growing interest in identifying suitable routes for the disposal of pharmaceutical wastes. This study investigates the potential of matrix materials composed of recycled polyethylene/polypropylene reclaimed from municipal solid wastes at immobilizing pharmaceutical solid wastes. Diclofenac (DF) drug product was embedded in boards of recycled plastic material, and leaching in water was assessed at various temperatures. DF concentrations were determined by high-performance liquid chromatography and revealed a maximum leachable fraction of 4% under accelerated conditions of 70°C, and less than 0.3% following 39 days of exposure at 20°C. The Ensemble Kalman Filter was employed to characterize the leaching behavior of DF. The filter verified the occurrence of leaching through diffusion, and was successful in predicting the leaching behavior of DF at 50°C and 70°C.

  10. Geochemical data package for the Hanford immobilized low-activity tank waste performance assessment (ILAW PA)

    International Nuclear Information System (INIS)

    DI Kaplan; RJ Serne

    2000-01-01

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as low-activity waste is to vitrify the liquid/slurry and place the solid product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is the Hanford Immobilized Low-Activity Tank Waste (ILAW) Performance Assessment (PA) activity. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require prediction of contaminant migration from the facilities. This migration is expected to occur primarily via the movement of water through the facilities, and the consequent transport of dissolved contaminants in the porewater of the vadose zone. Pacific Northwest National Laboratory assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of the geochemical properties of the materials comprising the disposal facility, the disturbed region around the facility, and the physically undisturbed sediments below the facility (including the vadose zone sediments and the aquifer sediments in the upper unconfined aquifer). The geochemical properties are expressed as parameters that quantify the adsorption of contaminants and the solubility constraints that might apply for those contaminants that may exceed solubility constraints. The common parameters used to quantify adsorption and solubility are the distribution coefficient (K d ) and the thermodynamic solubility product (K sp ), respectively. In this data package, the authors approximate the solubility of contaminants using a more simplified construct, called the

  11. Immobilization of Mn and NH4 (+)-N from electrolytic manganese residue waste.

    Science.gov (United States)

    Chen, Hongliang; Liu, Renlong; Liu, Zuohua; Shu, Jiancheng; Tao, Changyuan

    2016-06-01

    The objective of this work was the immobilization of soluble manganese (Mn) and ammonium nitrogen (NH4 (+)-N) leached from electrolytic manganese residue (EMR). Immobilization of Mn was investigated via carbonation using carbon dioxide (CO2) and alkaline additives. NH4 (+)-N immobilization was evaluated via struvite precipitation using magnesium and phosphate sources. Results indicated that the immobilization efficiency of Mn using CO2 and quicklime (CaO) was higher than using CO2 and sodium hydroxide (NaOH). This higher efficiency was likely due to the slower release of OH(-) during CaO hydrolysis. The immobilization efficiency of Mn was >99.99 % at the CaO:EMR mass ratio of 0.05:1 for 20-min reaction time. The struvite precipitation of NH4 (+)-N was conducted in the carbonated EMR slurry and the immobilization efficiency was 89 % using MgCl2 · 6H2O + Na3PO4 · 12H2O at the Mg:P:N molar ratio of 1.5:1.5:1 for 90-min reaction time. A leaching test showed that the concentrations of Mn and NH4 (+)-N in the filtrate of the treated EMR were 0.2 and 9 mg/L, respectively. The combined immobilization of Mn and NH4 (+)-N was an effective pretreatment method in the harmless treatment of the EMR.

  12. Production of immobilized cellulase enzyme by some microorganisms from the rice straw agro-waste using γ-irradiation

    International Nuclear Information System (INIS)

    Mohamed, M.A.Z.

    2014-01-01

    Studies were carried out using 14 fungal cultures screened for their ability to produce cellulase enzymes. A .hortai was selected for the present research as a potent cellulase producer. Cultural and nutritional factors affecting cellulase production were also investigated in order to optimize the fermentation conditions for the maximization of production. The obtained results revealed that, the maximum cellulase production (0.23 U/ml) was achieved after 96 h in a liquid medium (Ph 7.0) inoculated with 10% v/v inoculum size, at temperature 37 ºC, containing (gL -1 ) CMC, 5.0; yeast extract, 0.1; (NH 4 ) 2 SO 4 , 0.5; KH 2 PO 4 , 10.0; MgSO 4 .7H 2 O, 0.1 and NaCl, 0.2. The activity remained almost stable between ph 6.0 and 7.0. The highest cellulase activity (1.18 U/ml) was obtained at a lactose concentration of (5.0 gL -1 ). Partial purification of the crude cellulase by ammonium sulphate 70% saturation showed the highest specific enzyme activity and purification fold (2.3 U/mg protein and 2.12 fold, respectively). Different carriers and methods were used to select the suitable one for cellulase immobilization. Poly (acrylamide-co-acrylic acid) prepared by diazotization method increase S.E.A and the amount of immobilized enzyme to be (2.3 U/mg protein and 2.8 mg), respectively. The immobilized cellulase shows better operational stability, including wider ph and thermal ranges. The immobilized cellulase remained fully active up to 60°C. The kinetic parameters Km and Vmax were determined. The increase of the apparent Km after immobilization clearly indicates an apparent lower affinity of the immobilized enzyme for its substrate than the free enzyme. The resulting immobilized cellulase exhibited good reusability on degradation of rice straw agricultural wastes and also show good storage stability, that it lost only 17 % of its initial activity after 6 weeks.

  13. One step bioconversion of waste precious metals into Serratia biofilm-immobilized catalyst for Cr(VI) reduction.

    Science.gov (United States)

    Yong, P; Liu, W; Zhang, Z; Beauregard, D; Johns, M L; Macaskie, L E

    2015-11-01

    For reduction of Cr(VI) the Pd-catalyst is excellent but costly. The objectives were to prove the robustness of a Serratia biofilm as a support for biogenic Pd-nanoparticles and to fabricate effective catalyst from precious metal waste. Nanoparticles (NPs) of palladium were immobilized on polyurethane reticulated foam and polypropylene supports via adhesive biofilm of a Serratia sp. The biofilm adhesion and cohesion strength were unaffected by palladization and catalytic biofilm integrity was also shown by magnetic resonance imaging. Biofilm-Pd and mixed precious metals on biofilm (biofilm-PM) reduced 5 mM Cr(VI) to Cr(III) when immobilized in a flow-through column reactor, at respective flow rates of 9 and 6 ml/h. The lower activity of the latter was attributed to fewer, larger, metal deposits on the bacteria. Activity was lost in each case at pH 7 but was restored by washing with 5 mM citrate solution or by exposure of columns to solution at pH 2, suggesting fouling by Cr(III) hydroxide product at neutral pH. A 'one pot' conversion of precious metal waste into new catalyst for waste decontamination was shown in a continuous flow system based on the use of Serratia biofilm to manufacture and support catalytic Pd-nanoparticles.

  14. Experience in industrial operation of the plant for immobilizing radioactive wastes in thermosetting resins at the Ardennes Nuclear Power Station

    International Nuclear Information System (INIS)

    Haller, P.; Romestain, P.; Bruant, J.P.

    1983-01-01

    The French Atomic Energy Commission (CEA) has developed, at the Grenoble Centre for Nuclear Studies, a procedure for immobilizing low- and intermediate-level wastes in thermosetting resins of the polyester or epoxy types. To demonstrate feasibility on an industrial scale, a pilot plant has been set up at the effluent treatment station of the Ardennes Franco-Belgium Nuclear Power Station (SENA), which is a 305 MW(e) PWR type. Assembly work began in January 1979. After a period devoted to final adjustments and operation with inactive products, conditioning of active products began in January 1981. In the paper, the methods of conditioning the three types of waste (evaporation concentrates, ion exchange resins and filter cartridges) are described, experience of the start-up and operation of the plant is reported and the principal results of coating characterization tests are given. The results of tests on active and inactive products show that the characteristics of the materials obtained on an industrial scale match those of laboratory products and confirm their high quality with regard to mechanical behaviour, fire resistance, homogeneity and low-leachability. Industrial experience and economic comparisons show that the process of immobilizing waste from nuclear power stations in thermosetting resins offers an extremely interesting alternative to classical methods of conditioning. (author)

  15. Selective cesium removal from radioactive liquid waste by crown ether immobilized new class conjugate adsorbent

    OpenAIRE

    Awual, M. R.; 矢板 毅; 田口 富嗣; 塩飽 秀啓; 鈴木 伸一; 岡本 芳浩

    2014-01-01

    Conjugate materials can provide chemical functionality, enabling an assembly of the ligand complexation ability to metal ions that are important for applications, such as separation and removal devices. In this study, we developed ligand immobilized conjugate adsorbent for selective cesium (Cs) removal from wastewater. The adsorbent was synthesized by direct immobilization of DB24C8 onto inorganic mesoporous silica. The obtained results revealed that adsorbent had higher selectivity towards C...

  16. On immobilization of high-level waste in an Y–Al garnet-based cermet matrix in SHS conditions

    OpenAIRE

    Konovalov, E.E.; Lastov, A.I.; Nerozin, N.A.

    2015-01-01

    A method of high-level waste (HLW) radionuclide immobilization in a long-life matrix based on Y–Al garnet, a material highly chemically resistant to natural environments, has been developed for the ultimate HLW isolation from the environment. Model systems containing Ce, Nd, Sm, Zr, Mo, 238U, and 241Am were used in the study as simulators of HLW radionuclides. An energy-saving technology of self-propagating high-temperature synthesis (SHS) was employed to synthesize the matrix material with f...

  17. A process for the treatment of olive mill waste waters by immobilized cells.

    Directory of Open Access Journals (Sweden)

    ElYachioui, M.

    2005-06-01

    Full Text Available Mould strains were immobilized on sawdust from woods as a solid material for the treatment of Olive Mill Waste (OMW waters. Assays were carried out in flasks. The treatment process was monitored by physico-chemical determinations including pH, polyphenols and COD, which were followed up during the incubation time. In parallel the chemical inhibitory activity of OMW was confirmed biologically by the determination of some microorganisms in the medium including the plate count, yeasts and lactic acid bacteria. Results indicated that the polyphenol degradation level was 87 %. The COD was also reduced by 60 %. The pH of the effluent increased from 4.5 to 6.6. The microbial profiles showed their best growth during the treatment period indicating a removal of the inhibitory activities from the OMW waters. The growth patterns of all microorganism groups were similar and could reach high levels in the effluent.Cepas de moho fueron inmovilizadas sobre serrín de madera como material sólido para el tratamiento de aguas residuales de un molino de aceituna (OMW. Los ensayos se realizaron en matraces. El proceso de tratamiento se monitorizó mediante determinaciones físico-químicas incluyendo pH, polifenoles y DQO, que también se analizaron durante el tiempo de incubación. En paralelo, la actividad inhibidora química de las OMW se confirma biológicamente mediante su efecto sobre algunos microorganismos incluyendo levaduras y bactérias ácido lácticas. Los resultados indicaron que los polifenoles se degradan hasta un nivel del 87 %. La DQO se redujo también al 60 %. El pH del efluente aumentó de 4.5 a 6.6. Los perfiles microbiológicos mostraron un mejor crecimiento a medida que avanzaba el tratamiento indicando una supresión de las actividades inhibidoras de las aguas (OMW. El comportamiento del crecimiento de todos los grupos de microorganismos fue similar y puede alcanzar altos niveles en el efluente

  18. Synthesis of organic liquids/geo-polymer composites for the immobilization of nuclear wastes

    International Nuclear Information System (INIS)

    Cantarel, Vincent

    2016-01-01

    This work is included in the management of radioactive organic liquids research field. The process is based on an emulsification of organic liquid in an alkali silicate solution allowing the synthesis of a geo-polymer matrix. The first part of this work consists in carrying out a screening on different organic liquids. A model system representative of the various oils and a geo-polymer reference formulation are then defined. The second part deals with the structuration of the organic liquid/geo-polymer structuration, from the mixture of the reactants to the final material. It aims at determining the phenomena allowing the synthesis of a homogeneous composite. The last two parts aim at characterizing the composite by studying its structure (chemical structure, porosity of the geo-polymer and dispersion of the oil) and its properties with respect to the application to the immobilization of radioactive waste. Unlike calcium silicate-based cementitious matrices, the structure of the geo-polymer is not affected by the chemical nature of the organic liquids. Only acid oils inhibit or slow down the geo-polymerization reaction. In order to obtain a homogeneous material, the presence of surfactant molecules is necessary. The emulsion stabilization mechanism at the base of the process is relying on a synergy between the surfactant molecules and the aluminosilicate particles present in the geo-polymer paste. The kinetics (chemical and mechanical) of the geo-polymerization are not impacted by the presence of oil or surfactants. Only an increase in the viscoelastic moduli and the elastic character of the pastes can be observed. This difference in rheological behavior is mainly due to the presence of surfactant. The structure of the matrix is identical to that of a pure geo-polymer of the same formulation. The organic liquid is dispersed in spherical inclusions whose radius is between 5 and 15 μm. These droplets are separated from each other, and from the environment by the

  19. New aluminium-rich alkali slag matrix with clay minerals for immobilizing simulated radioactive Sr and Cs waste

    International Nuclear Information System (INIS)

    Qian Guangren; Sun, Darren Delai; Tay, Joo Hwa

    2001-01-01

    A new aluminium-rich alkali-activated slag matrix (M-AAS) with clay absorbents has been developed for immobilization of simulated radioactive Sr or Cs waste by introducing metakaolin, natural zeolite and NaOH-treated attapulgite clay minerals into alkali-activated slag matrix (AAS). The results revealed that the additions of metakaolin and clay absorbents into the cementitious matrixes would greatly enhance the distribution ratio, R d , of selective adsorption whether the matrix was OPC matrix or AAS matrix. The new immobilizing matrix M-AAS not only exhibited the strongest selective adsorption for both Sr and Cs ions, but also was characterized by lower porosity and small pore diameter so that it exhibited the lowest leaching rate. Hydration product analyses also demonstrated that (Na+Al)-substituted C-S-H(I) and self-generated zeolite were major hydration products in the M-AAS matrix, which provided this new immobilizing matrix with better selective adsorption on Sr and Cs and lower leaching rate

  20. The function of waste immobilization as a barrier against radionuclide release in the frame of an overall risk assessment

    International Nuclear Information System (INIS)

    Girardi, F.

    1983-01-01

    The general approach to safety assurance in disposal of radioactive waste is the 'multibarrier' or 'defence-in-depth' approach. Research has been mostly directed to characterize individual barriers in order to learn what performance may be expected from the barrier when utilized in a hypothetical safety system. The knowledge available is briefly reviewed, with special reference to ongoing activities carried out within the framework of the waste management programme of the European Communities. The interaction between different barriers in a geological disposal environment is not sufficiently well known to permit a full appreciation of the long-term behaviour of a barrier system. Examples of studies under way within the programme mentioned above are outlined briefly. On the basis of the information available, immobilization of high-level waste in borosilicate glass appears an acceptable barrier, well-integrated in the various barrier systems. Studies on the long-term properties of conditioned alpha-contaminated waste deserve a high priority in order to make a balanced evaluation of the long-term risks of the various waste types. (author)

  1. Application of solvlent change techniques to blended cements used to immobilize low-level radioactive liquid waste

    International Nuclear Information System (INIS)

    Kruger, A.A.

    1996-07-01

    The microstructures of hardened portland and blended cement pastes, including those being considered for use in immobilizing hazardous wastes, have a complex pore structure that changes with time. In solvent exchange, the pore structure is examined by immersing a saturated sample in a large volume of solvent that is miscible with the pore fluid. This paper reports the results of solvent replacement measurements on several blended cements mixed at a solution:solids ratio of 1.0 with alkaline solutions from the simulation of the off- gas treatment system in a vitrification facility treating low-level radioactive liquid wastes. The results show that these samples have a lower permeability than ordinary portland cement samples mixed at a water:solids ratio of 0.70, despite having a higher volume of porosity. The microstructure is changed by these alkaline solutions, and these changes have important consequences with regard to durability

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

    International Nuclear Information System (INIS)

    Supardi

    2000-01-01

    The experimental investigation of immobilization the low active wasteconcentration containing 2.44x10 -3 μCi/cc a great deal of NaNO 3 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/mm 2 . For the immobilization of the waste andcement with the optimum pH being used, yielded in the compressive strength of28.07 N/mm 2 . 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/mm 2 before irradiation, where as after irradiation thecompressive strength was 32.41 N/mm 2 . The optimum thermal test resultachieved was 250 o C with the compressive strength of 44.10 N/mm 2 . For theleaching test results after being cured for 91 days in the distilled watermedia was 0.47x10 -4 gcm -2 day -1 , while in the sea water was 0.66x10 -4 gcm -2 day -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

  3. Thermal immobilization of Cr, Cu and Zn of galvanizing wastes in the presence of clay and fly ash.

    Science.gov (United States)

    Singh, I B; Chaturvedi, K; Yegneswaran, A H

    2007-07-01

    In the present investigation thermal treatment of galvanizing waste with clay and fly ash has been carried out to immobilize Cr, Zn, Cu and other metals of the waste at temperature range 850 degrees C to 950 degrees C. Leaching of the metals from the waste and solidified product was analyzed using toxic characteristic leaching procedure (TCLP). Results indicated that the composition of waste and clay treatment temperature are the key factors in determining the stability of solidified product. After heating at 950 degrees C, the solidified specimens of 10% waste with clay have shown comparatively a high compressive strength and less water absorption. However, a decrease in compressive strength and increase in water absorption were noticed after addition of 15% of waste with clay. The leachability of all the metals present in the waste was found to reduce considerably with the increase of treatment temperature. In the case of Cr and Zn, their leachabilty was found at unacceptable levels from the treated product obtained after heating at 850 degrees C However, their leachability was reduced significantly within an acceptable level after treatment at 950 degrees C. The thermal treatment has shown an increase of re-oxidation trend of Cr (III) to Cr (VI) up to 900 degrees C of heating and this trend became almost zero after heating at 950 degrees C. Addition of fly ash did not show any improvement in strength, durability and leachability of metals from the thermally treated product. X-ray diffraction (XRD) analysis of the product confirmed the presence of mixed phases of oxides of toxic metals.

  4. Development Of A Macro-Batch Qualification Strategy For The Hanford Tank Waste Treatment And Immobilization Plant

    International Nuclear Information System (INIS)

    Herman, Connie C.

    2013-01-01

    The Savannah River National Laboratory (SRNL) has evaluated the existing waste feed qualification strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP) based on experience from the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) waste qualification program. The current waste qualification programs for each of the sites are discussed in the report to provide a baseline for comparison. Recommendations on strategies are then provided that could be implemented at Hanford based on the successful Macrobatch qualification strategy utilized at SRS to reduce the risk of processing upsets or the production of a staged waste campaign that does not meet the processing requirements of the WTP. Considerations included the baseline WTP process, as well as options involving Direct High Level Waste (HLW) and Low Activity Waste (LAW) processing, and the potential use of a Tank Waste Characterization and Staging Facility (TWCSF). The main objectives of the Hanford waste feed qualification program are to demonstrate compliance with the Waste Acceptance Criteria (WAC), determine waste processability, and demonstrate unit operations at a laboratory scale. Risks to acceptability and successful implementation of this program, as compared to the DWPF Macro-Batch qualification strategy, include: Limitations of mixing/blending capability of the Hanford Tank Farm; The complexity of unit operations (i.e., multiple chemical and mechanical separations processes) involved in the WTP pretreatment qualification process; The need to account for effects of blending of LAW and HLW streams, as well as a recycle stream, within the PT unit operations; and The reliance on only a single set of unit operations demonstrations with the radioactive qualification sample. This later limitation is further complicated because of the 180-day completion requirement for all of the necessary waste feed qualification steps. The primary recommendations/changes include the

  5. Recovery of uranium from low uranium concentration waste water using collagen fiber immobilized bayberry tannin

    International Nuclear Information System (INIS)

    Wu Yun; Long Xianming; Zhao Ning; Liao Pinxue

    2012-01-01

    Tannin, extracted from plants, is a kind of natural polyphenol, which is able to chelate with various metal ions and also exhibits selectivity in some extent. The collagen fiber immobilized bayberry tannin was prepared by the immobilization of bayberry tannin onto collagen fiber through the Mannich reaction. Experiment of the adsorption of U from U containing wastewater by using collagen fiber immobilized bayberry tannin suggested that the pH increase of U containing wastewater can promote the adsorption of U onto the adsorbent. When the pH was 4.5 and the initial concentration of U was 300.0 mg/L, the adsorption capacity of U reached the maximum of 52 mg/g while the other impurity metal ions were less than 16.0 mg/g, thus exhibiting excellent selectivity. The treatment of wastewater can be optimized by changing the U concentration, inlet rate of wastewater, and the ratio of column height/diameter etc. In addition. the adsorbed U can be desorbed using 0.1 mol/L HNO 3 solution when the column was saturated, the column can also be re used for the treatment of U containing wastewater after the column is washed by deionized water, collagen fiber immobilized bayberry tannin exhibit selectivity, high adsorption capacity, good reusability when adsorbed U. (authors)

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

    International Nuclear Information System (INIS)

    Chales Suarez, G.; Castillo Gomez, R.

    1997-01-01

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

  7. Scenarios for the Hanford Immobilized Low-Activity Waste (ILAW) performance assessment

    International Nuclear Information System (INIS)

    MANN, F.M.

    1999-01-01

    Scenarios describing representative exposure cases associated with the disposal of low activity waste from the Hanford Waste Tanks have been defined. These scenarios are based on guidance from the Department of Energy, the U.S. Nuclear Regulatory Commission, and previous Hanford waste disposal performance assessments

  8. Performance objectives for the Hanford Immobilized Low-Activity Waste (ILAW) performance assessment

    International Nuclear Information System (INIS)

    MANN, F.M.

    1999-01-01

    Performance objectives for the disposal of low activity waste from Hanford Waste Tanks have been developed. These objectives have been based on DOE requirements, programmatic requirements, and public involvement. The DOE requirements include regulations that direct the performance assessment and are cited within the Radioactive Waste Management Order (DOE Order 435.1). Performance objectives for other DOE complex performance assessments have been included

  9. Development of a new generation of waste form for entrapment and immobilization of highly volatile and soluble radionuclides.

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, Mark Andrew; Bencoe, Denise Nora; Brinker, C. Jeffrey; Murphy, Andrew Wilson; Holt, Kathleen Caroline; Turnham, Rigney; Kruichak, Jessica Nicole; Tellez, Hernesto; Miller, Andy; Xiong, Yongliang; Pohl, Phillip Isabio; Ockwig, Nathan W.; Wang, Yifeng; Gao, Huizhen

    2010-09-01

    The United States is now re-assessing its nuclear waste disposal policy and re-evaluating the option of moving away from the current once-through open fuel cycle to a closed fuel cycle. In a closed fuel cycle, used fuels will be reprocessed and useful components such as uranium or transuranics will be recovered for reuse. During this process, a variety of waste streams will be generated. Immobilizing these waste streams into appropriate waste forms for either interim storage or long-term disposal is technically challenging. Highly volatile or soluble radionuclides such as iodine ({sup 129}I) and technetium ({sup 99}Tc) are particularly problematic, because both have long half-lives and can exist as gaseous or anionic species that are highly soluble and poorly sorbed by natural materials. Under the support of Sandia National Laboratories (SNL) Laboratory-Directed Research & Development (LDRD), we have developed a suite of inorganic nanocomposite materials (SNL-NCP) that can effectively entrap various radionuclides, especially for {sup 129}I and {sup 99}Tc. In particular, these materials have high sorption capabilities for iodine gas. After the sorption of radionuclides, these materials can be directly converted into nanostructured waste forms. This new generation of waste forms incorporates radionuclides as nano-scale inclusions in a host matrix and thus effectively relaxes the constraint of crystal structure on waste loadings. Therefore, the new waste forms have an unprecedented flexibility to accommodate a wide range of radionuclides with high waste loadings and low leaching rates. Specifically, we have developed a general route for synthesizing nanoporous metal oxides from inexpensive inorganic precursors. More than 300 materials have been synthesized and characterized with x-ray diffraction (XRD), BET surface area measurements, and transmission electron microscope (TEM). The sorption capabilities of the synthesized materials have been quantified by using stable

  10. Development of a new generation of waste form for entrapment and immobilization of highly volatile and soluble radionuclides

    International Nuclear Information System (INIS)

    Rodriguez, Mark Andrew; Bencoe, Denise Nora; Brinker, C. Jeffrey; Murphy, Andrew Wilson; Holt, Kathleen Caroline; Turnham, Rigney; Kruichak, Jessica Nicole; Tellez, Hernesto; Miller, Andy; Xiong, Yongliang; Pohl, Phillip Isabio; Ockwig, Nathan W.; Wang, Yifeng; Gao, Huizhen

    2010-01-01

    The United States is now re-assessing its nuclear waste disposal policy and re-evaluating the option of moving away from the current once-through open fuel cycle to a closed fuel cycle. In a closed fuel cycle, used fuels will be reprocessed and useful components such as uranium or transuranics will be recovered for reuse. During this process, a variety of waste streams will be generated. Immobilizing these waste streams into appropriate waste forms for either interim storage or long-term disposal is technically challenging. Highly volatile or soluble radionuclides such as iodine ( 129 I) and technetium ( 99 Tc) are particularly problematic, because both have long half-lives and can exist as gaseous or anionic species that are highly soluble and poorly sorbed by natural materials. Under the support of Sandia National Laboratories (SNL) Laboratory-Directed Research and Development (LDRD), we have developed a suite of inorganic nanocomposite materials (SNL-NCP) that can effectively entrap various radionuclides, especially for 129 I and 99 Tc. In particular, these materials have high sorption capabilities for iodine gas. After the sorption of radionuclides, these materials can be directly converted into nanostructured waste forms. This new generation of waste forms incorporates radionuclides as nano-scale inclusions in a host matrix and thus effectively relaxes the constraint of crystal structure on waste loadings. Therefore, the new waste forms have an unprecedented flexibility to accommodate a wide range of radionuclides with high waste loadings and low leaching rates. Specifically, we have developed a general route for synthesizing nanoporous metal oxides from inexpensive inorganic precursors. More than 300 materials have been synthesized and characterized with x-ray diffraction (XRD), BET surface area measurements, and transmission electron microscope (TEM). The sorption capabilities of the synthesized materials have been quantified by using stable isotopes I and

  11. Evaluation of forms for the immobilization of high-level and transuranic wastes

    International Nuclear Information System (INIS)

    Schuman, R.P.; Cox, N.D.; Gibson, G.W.; Kelsey, P.V. Jr.

    1982-08-01

    A figure-of-merit (FOM) analysis has been made of a number of waste forms for solidifying both defense and commercial high-level reprocessing waste (HLW) and transuranic (TRU) wastes. The evaluation includes iron-enriched basalt (IEB), a fusion-produced glass-ceramic, which has not been included in other assessments. For HLW, concrete receives the highest FOM, but may not meet regulatory requirements; IEB and glass are the best choices of the materials that should easily meet regulatory requirements. Concrete waste forms are the best choice for TRU wastes, with IEB a close contender. 116 references, 3 figures, 112 tables

  12. An optimal retrieval, processing, and blending strategy for immobilization of Hanford high-level tank waste

    International Nuclear Information System (INIS)

    Hoza, M.

    1996-01-01

    Hanford tank waste will be separated into high-level and low-level portions; each portion will then be vitrified (other waste forms are also being considered for low-level waste) to produce a stable glass form for disposal. Because of the wide variability in the tank waste compositions, blending is being considered as a way to reduce the number of distinct compositions that must be vitrified and to minimize the resultant volume of vitrified waste. Three years of computational glass formulation and blending studies have demonstrated that blending of the high-level waste before vitrification can reduce the volume of high-level waste glass required by as much as 50 percent. This level of reduction would be obtained if all the high-level waste were blended together (Total Blend) prior to vitrification, requiring the retrieval and pretreatment of all tank waste before high-level vitrification was started. This paper will present an overall processing strategy that should be able to match the blending performance of the Total Blend and be more logistically feasible. The strategy includes retrieving, pretreating, blending and vitrifying Hanford tank waste. This strategy utilizes blending both before and after pretreatment. Similar wastes are blended before pretreatment, so as not to dilute species targeted for removal. The high-level portions of these pretreated early blends are then selectively blended to produce a small number of high-level vitrification feed streams

  13. In-situ vitrification: a large-scale prototype for immobilizing radioactively contaminated waste

    International Nuclear Information System (INIS)

    Carter, J.G.; Buelt, J.L.

    1986-03-01

    Pacific Northwest Laboratory is developing the technology of in situ vitrification, a thermal treatment process for immobilizing radioactively contaminated soil. A permanent remedial action, the process incorporates radionuclides into a glass and crystalline form. The transportable procss consists of an electrical power system to vitrify the soil, a hood to contain gaseous effluents, an off-gas treatment system and cooling system, and a process control station. Large-scale testing of the in situ vitrification process is currently underway

  14. Recommendations for computer code selection of a flow and transport code to be used in undisturbed vadose zone calculations for TWRS immobilized wastes environmental analyses

    International Nuclear Information System (INIS)

    VOOGD, J.A.

    1999-01-01

    An analysis of three software proposals is performed to recommend a computer code for immobilized low activity waste flow and transport modeling. The document uses criteria restablished in HNF-1839, ''Computer Code Selection Criteria for Flow and Transport Codes to be Used in Undisturbed Vadose Zone Calculation for TWRS Environmental Analyses'' as the basis for this analysis

  15. Effect of aluminum and silicon reactants and process parameters on glass-ceramic waste form characteristics for immobilization of high-level fluorinel-sodium calcined waste

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1993-06-01

    In this report, the effects of aluminum and silicon reactants, process soak time and the initial calcine particle size on glass-ceramic waste form characteristics for immobilization of the high-level fluorinel-sodium calcined waste stored at the Idaho Chemical Processing Plant (ICPP) are investigated. The waste form characteristics include density, total and normalized elemental leach rates, and microstructure. Glass-ceramic waste forms were prepared by hot isostatically pressing (HIPing) a pre-compacted mixture of pilot plant fluorinel-sodium calcine, Al, and Si metal powders at 1050 degrees C, 20,000 psi for 4 hours. One of the formulations with 2 wt % Al was HIPed for 4, 8, 16 and 24 hours at the same temperature and pressure. The calcine particle size range include as calcined particle size smaller than 600 μm (finer than -30 mesh, or 215 μm Mass Median Diameter, MMD) and 180 μm (finer than 80 mesh, or 49 μm MMD)

  16. Development and testing of a glass waste form for the immobilization of plutonium

    International Nuclear Information System (INIS)

    Chamberlain, D.B.; Hanchar, J.M.; Emery, J.W.; Hoh, J.C.; Wolf, S.F.; Finch, R.J.; Bates, J.K.; Ellison, A.J.G.; Dingwell, D.B.

    1996-01-01

    The United States has declared about 50 metric tons of weapons-grade Pu surplus to national security needs. The President has directed that this Pu be placed in a form that provides a high degree of proliferation resistance in which the surplus Pu is both unattractive and inaccessible for use by others [I]. Three alternatives are being evaluated for the disposal 2048 of this material: (1) use of the Pu as a fuel source for commercial reactors; (2) immobilization, where Pu is fixed in a glass or ceramic matrix that also contains or is surrounded by highly radioactive material; and (3) deep bore hole, where Pu is emplaced at depths of several kilometers. The immobilization alternative is being directed by the staff at Lawrence Livermore National Laboratory (LLNL). The staff at ANL are assisting by developing a glass for the immobilization of Pu and in the corrosion testing of glass and ceramic material prepared both at ANL and at other DOE laboratories. As part of this program, we have developed an ATS glass into which 5-7 wt percent Pu has been dissolved. The ATS glass was engineered to accommodate high Pu loading and to be durable under conditions likely to accelerate glass reactions in the geological environment during long-term storage

  17. Process Testing Results and Scaling for the Hanford Waste Treatment and Immobilization Plant (WTP) Pretreatment Engineering Platform - 10173

    International Nuclear Information System (INIS)

    Kurath, Dean E.; Daniel, Richard C.; Baldwin, David L.; Rapko, Brian M.; Barnes, Steven M.; Gilbert, Robert A.; Mahoney, Lenna A.; Huckaby, James L.

    2010-01-01

    The U.S. Department of Energy-Office of River Protections Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being designed and built to pretreat and then vitrify a large portion of the wastes in Hanfords 177 underground waste storage tanks at Richland, Washington. In support of this effort, engineering-scale tests at the Pretreatment Engineering Platform (PEP) have been completed to confirm the process design and provide improved projections of system capacity. The PEP is a 1/4.5-scale facility designed, constructed, and operated to test the integrated leaching and ultrafiltration processes being deployed at the WTP. The PEP replicates the WTP leaching processes with prototypic equipment and control strategies and non-prototypic ancillary equipment to support the core processing. The testing approach used a nonradioactive aqueous slurry simulant to demonstrate the unit operations of caustic and oxidative leaching, cross-flow ultrafiltration solids concentration, and solids washing. Parallel tests conducted at the laboratory scale with identical simulants provided results that allow scale-up factors to be developed between the laboratory and PEP performance. This paper presents the scale-up factors determined between the laboratory and engineering-scale results and presents arguments that extend these results to the full-scale process.

  18. Physical, Hydraulic, and Transport Properties of Sediments and Engineered Materials Associated with Hanford Immobilized Low-Activity Waste

    Energy Technology Data Exchange (ETDEWEB)

    Rockhold, Mark L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhang, Z. F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Meyer, Philip D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thomle, Jonathan N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-02-28

    Current plans for treatment and disposal of immobilized low-activity waste (ILAW) from Hanford’s underground waste storage tanks include vitrification and storage of the glass waste form in a nearsurface disposal facility. This Integrated Disposal Facility (IDF) is located in the 200 East Area of the Hanford Central Plateau. Performance assessment (PA) of the IDF requires numerical modeling of subsurface flow and reactive transport processes over very long periods (thousands of years). The models used to predict facility performance require parameters describing various physical, hydraulic, and transport properties. This report provides updated estimates of physical, hydraulic, and transport properties and parameters for both near- and far-field materials, intended for use in future IDF PA modeling efforts. Previous work on physical and hydraulic property characterization for earlier IDF PA analyses is reviewed and summarized. For near-field materials, portions of this document and parameter estimates are taken from an earlier data package. For far-field materials, a critical review is provided of methodologies used in previous data packages. Alternative methods are described and associated parameters are provided.

  19. Immobilization of INEL low-level radioactive wastes in ceramic containment materials

    International Nuclear Information System (INIS)

    Seymour, W.C.; Kelsey, P.V.

    1978-11-01

    INEL low-level radioactive wastes have an overall chemical composition that lends itself to self-containment in a ceramic-based material. Fewer chemical additives would be needed to process the wastes than to process high-level wastes or use a mixture containment method. The resulting forms of waste material could include a basalt-type glass or glass ceramic and a ceramic-type brick. Expected leach resistance is discussed in relationshp to data found in the literature for these materials and appears encouraging. An overview of possible processing steps for the ceramic materials is presented

  20. Fluidized bed steam reformed mineral waste form performance testing to support Hanford Supplemental Low Activity Waste Immobilization Technology Selection

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pierce, E. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bannochie, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Burket, P. R. [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); Crawford, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Herman, C. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [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); Nash, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Brown, C. F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qafoku, N. P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Neeway, J. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Valenta, M. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gill, G. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Swanberg, D. J. [Washington River Protection Solutions (WRPS), Richland, WA (United States); Robbins, R. A. [Washington River Protection Solutions (WRPS), Richland, WA (United States); Thompson, L. E. [Washington River Protection Solutions (WRPS), Richland, WA (United States)

    2015-10-01

    This report describes the benchscale testing with simulant and radioactive Hanford Tank Blends, mineral product characterization and testing, and monolith testing and characterization. These projects were funded by DOE EM-31 Technology Development & Deployment (TDD) Program Technical Task Plan WP-5.2.1-2010-001 and are entitled “Fluidized Bed Steam Reformer Low-Level Waste Form Qualification”, Inter-Entity Work Order (IEWO) M0SRV00054 with Washington River Protection Solutions (WRPS) entitled “Fluidized Bed Steam Reforming Treatability Studies Using Savannah River Site (SRS) Low Activity Waste and Hanford Low Activity Waste Tank Samples”, and IEWO M0SRV00080, “Fluidized Bed Steam Reforming Waste Form Qualification Testing Using SRS Low Activity Waste and Hanford Low Activity Waste Tank Samples”. This was a multi-organizational program that included Savannah River National Laboratory (SRNL), THOR® Treatment Technologies (TTT), Pacific Northwest National Laboratory (PNNL), Oak Ridge National Laboratory (ORNL), Office of River Protection (ORP), and Washington River Protection Solutions (WRPS). The SRNL testing of the non-radioactive pilot-scale Fluidized Bed Steam Reformer (FBSR) products made by TTT, subsequent SRNL monolith formulation and testing and studies of these products, and SRNL Waste Treatment Plant Secondary Waste (WTP-SW) radioactive campaign were funded by DOE Advanced Remediation Technologies (ART) Phase 2 Project in connection with a Work-For-Others (WFO) between SRNL and TTT.

  1. Photocatalytic degradation of an organophosphorus pesticide from agricultural waste by immobilized TiO2 under solar radiation

    Directory of Open Access Journals (Sweden)

    Marcia Regina Assalin

    2016-11-01

    Full Text Available This paper describes solar heterogeneous photocatalysis using immobilized TiO2 applied in the treatment of agricultural waste resulting from the application of commercial formulations of methyl parathion. The disappearance of the insecticide, as well as the formation of its metabolite, was monitored by high-performance liquid chromatography-tandem mass spectrometry (LC-MS/MS, while mineralization efficiency was monitored by measurements of total organic carbon (TOC. Toxicity studies were performed using the microcrustacean Artemia salina. The TOC removal efficiency by photocatalytic process was 48.5%. After 45 minutes of treatment, the removal efficiency of methyl parathion was 90%, being completely mineralized at the end of treatment. The formation and removal of the metabolite methyl paraoxon was observed during the photocatalytic process. The photocatalytic treatment resulted in increased microcrustacean mobility, indicating a reduction of acute toxicity.

  2. Evaluation and review of alternative waste forms for immobilization of high level radioactive wastes. Report number 3

    International Nuclear Information System (INIS)

    1981-01-01

    A discussion of the relative strengths and weaknesses of the alternative forms and recommendations for future program directions are presented in the body of this report. In addition to the relative ranking, the Peer Review Panel makes the following observations and recommendations: (1) Differences in overall performance of most of the uncoated waste forms are relatively small when compared under approximately equivalent conditions. (2) The increased scientific basis for this class of waste forms has not yet been sufficient to achieve reliably large improvements in waste form performance over the best borosilicate glasses. (3) The increased leach rates at elevated temperatures and the uncertainty regarding mechanisms of leaching under repository conditions continue to indicate that surface temperatures of waste canisters and especially any waste form-water interfaces should be restricted to less than 100 0 C, until more data is available to indicate otherwise. (4) Improvements are noteworthy, but there is still a need for adopting additional standardized tests, standard reference materials, common units and standardized methods of data presentation in the nuclear waste program. (5) Comparative data on leach rates in waters equilibrated with candidate rocks and potential geologic environments are almost non-existent and are essential to establish relevant long term extrapolation of waste form performance.(6) Understanding radiation damage effects on the microstructure and leaching mechanisms of polycrystalline ceramics is still insufficient to judge long term reliability of this class of waste forms. (7) More extensive data on rates and mechanisms of leaching of all waste forms under radiation and repository conditions are needed. (8) Additional studies of fundamental mechanisms controlling long term reliability of glass and alternative waste forms are strongly encouraged

  3. Immobilization of metal wastes by reaction with H2S in anoxic basins. Concept and Elaboration

    NARCIS (Netherlands)

    Schuiling, R.D.

    2013-01-01

    Metal wastes are produced in large quantities by a number of industries. Their disposal in isolated waste deposits is certain to cause many subsequent problems, because every material will sooner or later return to the geochemical cycle. The sealing of disposal sites usually starts to

  4. Assessment of processes, facilities, and costs for alternative solid forms for immobilization of SRP defense waste

    International Nuclear Information System (INIS)

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

    1982-03-01

    A quantitative merit evaluation which assesses the relative difficulty of remote processing of Savannah River Plant high-level wastes for seven alternative waste forms is presented. The reference borosilicate glass process is rated as the simplest, followed by FUETAP concrete. The other processes evaluated in order of increasing complexity were: glass marbles in a lead matrix, high-silica glass, crystalline ceramic (Synroc-D and tailored ceramic), and coated ceramic particles. Cost appraisals are summarized for the borosilicate glass, high-silica glass, and ceramic waste form processing facilities

  5. Selective cesium removal from radioactive liquid waste by crown ether immobilized new class conjugate adsorbent.

    Science.gov (United States)

    Awual, Md Rabiul; Yaita, Tsuyoshi; Taguchi, Tomitsugu; Shiwaku, Hideaki; Suzuki, Shinichi; Okamoto, Yoshihiro

    2014-08-15

    Conjugate materials can provide chemical functionality, enabling an assembly of the ligand complexation ability to metal ions that are important for applications, such as separation and removal devices. In this study, we developed ligand immobilized conjugate adsorbent for selective cesium (Cs) removal from wastewater. The adsorbent was synthesized by direct immobilization of dibenzo-24-crown-8 ether onto inorganic mesoporous silica. The effective parameters such as solution pH, contact time, initial Cs concentration and ionic strength of Na and K ion concentrations were evaluated and optimized systematically. This adsorbent was exhibited the high surface area-to-volume ratios and uniformly shaped pores in case cavities, and its active sites kept open functionality to taking up Cs. The obtained results revealed that adsorbent had higher selectivity toward Cs even in the presence of a high concentration of Na and K and this is probably due to the Cs-π interaction of the benzene ring. The proposed adsorbent was successfully applied for radioactive Cs removal to be used as the potential candidate in Fukushima nuclear wastewater treatment. The adsorbed Cs was eluted with suitable eluent and simultaneously regenerated into the initial form for the next removal operation after rinsing with water. The adsorbent retained functionality despite several cycles during sorption-elution-regeneration operations. Copyright © 2014 Elsevier B.V. All rights reserved.

  6. Deproteination of shrimp shell wastes using immobilized marine associated pseudomonad Amet1776

    Digital Repository Service at National Institute of Oceanography (India)

    Bhagat, J.; Venkatramani, M.; Hussain, A. J.; Jayaprakashvel, M.

    in abundance in India in the last few years. The conventional demineralization, deproteination and decoloration method of extraction of chitin from crustacean waste is costly and causes environmental problems. In this study bioconversion of chitinous material...

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-15

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  10. A review of phase separation in borosilicate glasses, with reference to nuclear fuel waste immobilization

    International Nuclear Information System (INIS)

    Taylor, P.

    1990-08-01

    This report reviews information on miscibility limits in borosilicate glass-forming systems. It includes both a literature survey and an account of experimental work performed within the Canadian Nuclear Fuel Waste Management Program. Emphasis is placed on the measurement and depiction of miscibility limits in multicomponent (mainly quaternary) systems, and the effects of individual components on the occurrence of phase separation. The behaviour of the multicomponent system is related to that of simpler (binary and ternary) glass systems. The possible occurrence of phase separation, as well as its avoidance, during processing of nuclear waste glasses is discussed

  11. DEVELOPMENT OF CRYSTALLINE CERAMICS FOR IMMOBILIZATION OF ADVANCED FUEL CYCLE REPROCESSING WASTES

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Brinkman, K.

    2011-09-22

    The Savannah River National Laboratory (SRNL) is developing crystalline ceramic waste forms to incorporate CS/LN/TM high Mo waste streams consisting of perovskite, hollandite, pyrochlore, zirconolite, and powellite phase assemblages. Simple raw materials, including Al{sub 2}O{sub 3}, CaO, and TiO{sub 2} were combined with simulated waste components to produce multiphase crystalline ceramics. Fiscal Year 2011 (FY11) activities included (i) expanding the compositional range by varying waste loading and fabrication of compositions rich in TiO{sub 2}, (ii) exploring the processing parameters of ceramics produced by the melt and crystallize process, (iii) synthesis and characterization of select individual phases of powellite and hollandite that are the target hosts for radionuclides of Mo, Cs, and Rb, and (iv) evaluating the durability and radiation stability of single and multi-phase ceramic waste forms. Two fabrication methods, including melting and crystallizing, and pressing and sintering, were used with the intent of studying phase evolution under various sintering conditions. An analysis of the XRD and SEM/EDS results indicates that the targeted crystalline phases of the FY11 compositions consisting of pyrochlore, perovskite, hollandite, zirconolite, and powellite were formed by both press and sinter and melt and crystallize processing methods. An evaluation of crystalline phase formation versus melt processing conditions revealed that hollandite, perovskite, zirconolite, and residual TiO{sub 2} phases formed regardless of cooling rate, demonstrating the robust nature of this process for crystalline phase development. The multiphase ceramic composition CSLNTM-06 demonstrated good resistance to proton beam irradiation. Electron irradiation studies on the single phase CaMoO{sub 4} (a component of the multiphase waste form) suggested that this material exhibits stability to 1000 years at anticipated self-irradiation doses (2 x 10{sup 10}-2 x 10{sup 11} Gy), but that

  12. Waste Treatment And Immobilization Plant U. S. Department Of Energy Office Of River Protection Submerged Bed Scrubber Condensate Disposition Project - Abstract no. 13460

    International Nuclear Information System (INIS)

    Yanochko, Ronald M; Corcoran, Connie

    2012-01-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) will generate an off-gas treatment system secondary liquid waste stream [submerged bed scrubber (SBS) condensate], which is currently planned for recycle back to the WTP Low Activity Waste (LAW) melter. This SBS condensate waste stream is high in Tc-99, which is not efficiently captured in the vitrified glass matrix. A pre-conceptual engineering study was prepared in fiscal year 2012 to evaluate alternate flow paths for melter off-gas secondary liquid waste generated by the WTP LAW facility. This study evaluated alternatives for direct off-site disposal of this SBS without pre-treatment, which mitigates potential issues associated with recycling

  13. Waste Treatment And Immobilization Plant U. S. Department Of Energy Office Of River Protection Submerged Bed Scrubber Condensate Disposition Project - Abstract # 13460

    Energy Technology Data Exchange (ETDEWEB)

    Yanochko, Ronald M [Washington River Protection Solutions, Richland, WA (United States); Corcoran, Connie [AEM Consulting, LLC, Richland, WA (United States)

    2012-11-15

    The Hanford Waste Treatment and Immobilization Plant (WTP) will generate an off-gas treatment system secondary liquid waste stream [submerged bed scrubber (SBS) condensate], which is currently planned for recycle back to the WTP Low Activity Waste (LAW) melter. This SBS condensate waste stream is high in Tc-99, which is not efficiently captured in the vitrified glass matrix. A pre-conceptual engineering study was prepared in fiscal year 2012 to evaluate alternate flow paths for melter off-gas secondary liquid waste generated by the WTP LAW facility. This study evaluated alternatives for direct off-site disposal of this SBS without pre-treatment, which mitigates potential issues associated with recycling.

  14. Performance objectives for the Hanford immobilized low-activity waste (ILAW) performance assessment

    International Nuclear Information System (INIS)

    MANN, F.M.

    1999-01-01

    Before low-level waste may be disposed of, a performance assessment must be written and then approved by the DOE (DOE 1988a, DOE 1999a). The performance assessment is to determine whether ''reasonable assurance'' exists that the performance objectives of the disposal facility will be met. The DOE requirements for waste disposal (DOE 1988a, DOE 1999a) require (see Appendix B): The protection of public health and safety; and The protection of the environment. Although quantitative limits are sometimes stated (for example, the all-pathways exposure limit is 25 mredyear), usually the requirements are stated in a general nature. Quantitative limits were established by: investigating all potentially applicable regulations as well as interpretations of the review panels which DOE has established to review performance assessments; interacting with program management to establish the additional requirements of the program; and interacting with the public (i.e., the Hanford Advisory Board members; as well as affected Tribal Governments) to understand the values of residents in the Pacific Northwest. Because of space considerations, not all radionuclides and dangerous chemicals are listed in this document. The radionuclides listed here are those which were explicitly treated in the ILAW PA (Mann 1998). The dangerous chemicals listed here are those most often detected in Hanford tank waste as documented in the Regulatory Data Quality Objectives Supporting Tank Waste Remediation System Privatization Project (Wiemers 1998)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  16. Polyphase ceramic and glass-ceramic forms for immobilizing ICPP high-level nuclear waste

    International Nuclear Information System (INIS)

    Harker, A.B.; Flintoff, J.F.

    1984-01-01

    Polyphase ceramic and glass-ceramic forms have been consolidated from simulated Idaho Chemical Processing Plant wastes by hot isostatic pressing calcined waste and chemical additives by 1000 0 C or less. The ceramic forms can contain over 70 wt% waste with densities ranging from 3.5 to 3.85 g/cm 3 , depending upon the formulation. Major phases are CaF 2 , CaZrTi 207 , CaTiO 3 , monoclinic ZrO 2 , and amorphous intergranular material. The relative fraction of the phases is a function of the chemical additives (TiO 2 , CaO, and SiO 2 ) and consolidation temperature. Zirconolite, the major actinide host, makes the ceramic forms extremely leach resistant for the actinide simulant U 238 . The amorphous phase controls the leach performance for Sr and Cs which is improved by the addition of SiO 2 . Glass-ceramic forms were also consolidated by HIP at waste loadings of 30 to 70 wt% with densities of 2.73 to 3.1 g/cm 3 using Exxon 127 borosilicate glass frit. The glass-ceramic forms contain crystalline CaF 2 , Al 203 , and ZrSi 04 (zircon) in a glass matrix. Natural mineral zircon is a stable host for 4+ valent actinides. 17 references, 3 figures, 5 tables

  17. Glass Formulation For The Hanford Tank Waste Treatment And Immobilization Plant (WTP)

    International Nuclear Information System (INIS)

    Kruger, A.A.; Jain, V.

    2009-01-01

    A computational method for formulating Hanford HLW glasses was developed that is based on empirical glass composition-property models, accounts for all associated uncertainties, and can be solved in Excel R in minutes. Calculations for all waste form processing and compliance requirements included. Limited experimental validation performed.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  19. GLASS FORMULATION FOR THE HANFORD TANK WASTE TREATMENT AND IMMOBILIZATION PLANT (WTP)

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; VIENNA JD; KIM DS; JAIN V

    2009-05-27

    A computational method for formulating Hanford HLW glasses was developed that is based on empirical glass composition-property models, accounts for all associated uncertainties, and can be solved in Excel{sup R} in minutes. Calculations for all waste form processing and compliance requirements included. Limited experimental validation performed.

  20. One-step synthesis of high-yield biodiesel from waste cooking oils by a novel and highly methanol-tolerant immobilized lipase.

    Science.gov (United States)

    Wang, Xiumei; Qin, Xiaoli; Li, Daoming; Yang, Bo; Wang, Yonghua

    2017-07-01

    This study reported a novel immobilized MAS1 lipase from marine Streptomyces sp. strain W007 for synthesizing high-yield biodiesel from waste cooking oils (WCO) with one-step addition of methanol in a solvent-free system. Immobilized MAS1 lipase was selected for the transesterification reactions with one-step addition of methanol due to its much more higher biodiesel yield (89.50%) when compared with the other three commercial immobilized lipases (biodiesel yield (95.45%) was acquired with one-step addition of methanol under the optimized conditions. Moreover, it was observed that immobilized MAS1 lipase retained approximately 70% of its initial activity after being used for four batch cycles. Finally, the obtained biodiesel was further characterized using FT-IR, 1 H and 13 C NMR spectroscopy. These findings indicated that immobilized MAS1 lipase is a promising catalyst for biodiesel production from WCO with one-step addition of methanol under high methanol concentration. Copyright © 2017 Elsevier Ltd. All rights reserved.

  1. Bentonite-Clay Waste Form for the Immobilization of Cesium and Strontium from Fuel Processing Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Kaminski, Michael D. [Argonne National Lab. (ANL), Argonne, IL (United States); Mertz, Carol J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-01-01

    The physical properties of a surrogate waste form containing cesium, strontium, rubidium, and barium sintered into bentonite clay were evaluated for several simulant feed streams: chlorinated cobalt dicarbollide/polyethylene glycol (CCD-PEG) strip solution, nitrate salt, and chloride salt feeds. We sintered bentonite clay samples with a loading of 30 mass% of cesium, strontium, rubidium, and barium to a density of approximately 3 g/cm3. Sintering temperatures of up to 1000°C did not result in volatility of cesium. Instead, there was an increase in crystallinity of the waste form upon sintering to 1000ºC for chloride- and nitrate-salt loaded clays. The nitrate salt feed produced various cesium pollucite phases, while the chloride salt feed did not produce these familiar phases. In fact, many of the x-ray diffraction peaks could not be matched to known phases. Assemblages of silicates were formed that incorporated the Sr, Rb, and Ba ions. Gas evolution during sintering to 1000°C was significant (35% weight loss for the CCD-PEG waste-loaded clay), with significant water being evolved at approximately 600°C.

  2. Supply and cost factors for metals in the Canadian nuclear fuel waste immobilization program

    International Nuclear Information System (INIS)

    McConnell, D.B.

    1982-11-01

    Estimates have been made of the demand for immobilization containers to accommodate the irradiated fuel bundles arising from Canadian nuclear generating stations to the year 2020. The resulting estimates for container shells and container-filling alloys were compared to estimates for Canadian and Western World production of the candiate metals. The results indicate that, among the container shell metals, supply difficulties might arise only for Grade 7 titanium. Among the filling metals, only lead-antimony alloy might present supply problems. Current cost figures for plate made of each shell metal, and bulk quantities of filling metals, were compared. Materials costs would be least for a supported shell of stainless steel, followed by copper, titanium alloys Grades 2, 12 and 7, and Inconel 625. Aluminum-silicon is the lowest-cost filling matrix, followed by zinc, lead, and lead-antimony. Container durability, vault conditions, groundwater composition and other factors may play an overriding role in the final selection of materials for container construction

  3. The structures and stability of media intended for the immobilization of high level radioactive waste

    International Nuclear Information System (INIS)

    Tempest, P.A.

    1979-05-01

    High level radioactive waste contains about 40 different elements and, in time, many of these elements are transformed by radioactive decay into different-sized atoms with new chemical properties. The suitability of ordered crystal structures and unordered glass structures as media for immobilising the waste elements is compared. The structural properties of a mixture of synthetic minerals (SYNROC) are described and the various minerals' ability to accommodate ions of different radii and charge assessed. Similary the unordered structure of glass is examined and the probability of the glass remaining non-crystalline during manufacture and storage taken into account. Alternative glassification technologies in the form of the French AVM continuous process and the UK HARVEST batch processes are described and compared, and their likely effect on the structural properties of the final solid glass block considered. (author)

  4. Research of ceramic matrix for a safe immobilization of radioactive sludge waste

    Science.gov (United States)

    Dorofeeva, Ludmila; Orekhov, Dmitry

    2018-03-01

    The research and improvement of the existing method for radioactive waste hardening by fixation in a ceramic matrix was carried out. For the samples covered with the sodium silicate and tested after the storage on the air the speed of a radionuclides leaching was determined. The properties of a clay ceramics and the optimum conditions of sintering were defined. The experimental data about the influence of a temperature mode sintering, water quantities, sludge and additives in the samples on their mechanical durability and a water resistance were obtained. The comparative analysis of the conducted research is aimed at improvement of the existing method of the hardening radioactive waste by inclusion in a ceramic matrix and reveals the advantages of the received results over analogs.

  5. An experimental study on Sodalite and SAP matrices for immobilization of spent chloride salt waste

    Science.gov (United States)

    Giacobbo, Francesca; Da Ros, Mirko; Macerata, Elena; Mariani, Mario; Giola, Marco; De Angelis, Giorgio; Capone, Mauro; Fedeli, Carlo

    2018-02-01

    In the frame of Generation IV reactors a renewed interest in pyro-processing of spent nuclear fuel is underway. Molten chloride salt waste arising from the recovering of uranium and plutonium through pyro-processing is one of the problematic wastes for direct application of vitrification or ceramization. In this work, Sodalite and SAP have been evaluated and compared as potential matrices for confinement of spent chloride salt waste coming from pyro-processing. To this aim Sodalite and SAP were synthesized both in pure form and mixed with different glass matrices, i.e. commercially available glass frit and borosilicate glass. The confining matrices were loaded with mixed chloride salts to study their retention capacities with respect to the elements of interest. The matrices were characterized and leached for contact times up to 150 days at room temperature and at 90 °C. SEM analyses were also performed in order to compare the matrix surface before and after leaching. Leaching results are discussed and compared in terms of normalized releases with similar results reported in literature. According to this comparative study the SAP matrix with glass frit binder resulted in the best matrix among the ones studied, with respect to retention capacities for both matrix and spent fuel elements.

  6. A review and discussion of candidate ceramics for immobilization of high-level fuel reprocessing wastes

    International Nuclear Information System (INIS)

    Hayward, P.J.

    1982-08-01

    This review discusses and attempts to evaluate 11 of the leading ceramic processes for hosting the high-level and high-level plus medium-level wastes which would arise from the reprocessing of used UO 2 , (Th,Pu)O 2 and (Th,U)O 2 fuels. The wasteform materials considered include glass ceramics, supercalcine ceramics, SYNROC ceramics, 'stuffed glass', titanate ceramics, cermets, clay ceramics, cement-based materials and multibarrier wasteforms. Although no attempt has been made to rank these candidates in order of superiority, the conclusion is drawn that, of the materials proposed so far, a glass ceramic appears to be best suited to the Canadian program, taking into account durability in the potential environment of a flooded vault, ability to withstand radiation and transmutation damage without serious loss of durability, ability to accommodate variable waste compositions, and ease of processing and quality control. This conclusion does not necessarily apply to other national waste management programs. However, many of the points raised might be included in any critical assessment of alternative wasteform materials

  7. Immobilization of simulated high-level radioactive waste in borosilicate glass: Pilot scale demonstrations

    International Nuclear Information System (INIS)

    Ritter, J.A.; Hutson, N.D.; Zamecnik, J.R.; Carter, J.T.

    1991-01-01

    The Integrated DWPF Melter System (IDMS), operated by the Savannah River Laboratory, is a pilot scale facility used in support of the start-up and operation of the Department of Energy's Defense Waste Processing Facility. The IDMS has successfully demonstrated, on an engineering scale (one-fifth), that simulated high level radioactive waste (HLW) sludge can be chemically treated with formic acid to adjust both its chemical and physical properties, and then blended with simulated precipitate hydrolysis aqueous (PHA) product and borosilicate glass frit to produce a melter feed which can be processed into a durable glass product. The simulated sludge, PHA and frit were blended, based on a product composition program, to optimize the loading of the waste glass as well as to minimize those components which can cause melter processing and/or glass durability problems. During all the IDMS demonstrations completed thus far, the melter feed and the resulting glass that has been produced met all the required specifications, which is very encouraging to future DWPF operations. The IDMS operations also demonstrated that the volatile components of the melter feed (e.g., mercury, nitrogen and carbon, and, to a lesser extent, chlorine, fluorine and sulfur) did not adversely affect the melter performance or the glass product

  8. Study of powellite-rich glass-ceramics for nuclear waste immobilization

    International Nuclear Information System (INIS)

    Taurines, T.

    2012-01-01

    MoO 3 is poorly soluble in borosilicate glasses which can lead to the crystallization of undesired phases when its concentration or the charge load (minor actinides and fission products concentration) is too high. Crystallization control is needed to guarantee good immobilization properties. We studied powellite-rich glass-ceramics obtained from a simplified nuclear glass in the system SiO 2 - B 2 O 3 - Na 2 O - CaO - Al 2 O 3 - MoO 3 - RE 2 O 3 (RE = Gd, Eu, Nd) by various heat treatments. Rare earth elements (REE) were added as minor actinides surrogates and as spectroscopic probes. The influence of MoO 3 and RE 2 O 3 content on powellite (CaMoO 4 ) crystallization was investigated. Various glass-ceramics (similar residual glass + powellite) were obtained with large crystal size distributions. Phase separation due to molybdenum occurs during quenching when [MoO 3 ] ≥ 2.5 mol%. We showed that increasing the rare earth content can suppress the phase separation due to molybdenum but it leads to spinodal decomposition of the residual glass. Furthermore, we studied the effects of parent glass complexifying and the insertion of Gd 3+ ions into the powellite structure. In order to understand the influence of microstructure on evolutions under β-irradiation, we studied point defects creation and structural changes. We showed that the damage induced by electronic excitations in the glass-ceramics is driven by the damage in the residual glass. (author) [fr

  9. Development and utilization of bitumen immobilizing process for radioactive waste materials; Razvoj i primena procesa imobilizacije radioaktivnog otpadnog materijala bitumenskim postupkom

    Energy Technology Data Exchange (ETDEWEB)

    Peric, A; Plecas, I; Kostadinovic, A [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

    1988-07-01

    Bitumenization is one of the relevant methods for solidification of radioactive waste materials in nuclear technology, today. Bitumenization of R.A. wastes can be obtained either with batch or continual process. In 'IBK', mixer-evaporator vessel for batch process, with continual feeder of process influents, was developed. From the several R.A. materials, that can be solidified into bitumen, ion-exchanged resins, used in NPP Krsko, were chosen and treated, at the various ranges of pH and T. Researches and experiments pointed and verified on use of Yugoslav produced bitumens, as a matrix media for immobilizing the solid R.A. (author)

  10. Concentration and immobilization of 137Cs from liquid radioactive waste using sorbents based on hydrated titanium and zirconium oxides

    Science.gov (United States)

    Voronina, A. V.; Noskova, A. Y.; Gritskevich, E. Y.; Mashkovtsev, M. A.; Semenishchev, V. S.

    2017-09-01

    The possibility of use of sorbents based on hydrated titanium and zirconium oxides (T-3A, T-35, NPF-HTD) for concentration and immobilization of 137Cs from liquid radioactive waste of various chemical composition (fresh water, seawater, solutions containing NaNO3, ammonium acetate, EDTA) was evaluated. It was shown that the NPF-HTD and T-35 sorbents separate 137Cs from fresh water and seawater with distribution coefficients as high as 6.2.104 and 6.1.104, 4.0.105 and 1.6.105 L kg-1 respectively; in 1 M ammonium acetate these values were 2.0.103 and 1.0.103 L kg-1. The NPF-HTD sorbent showed the highest selectivity for cesium in NaNO3 solution: cesium distribution coefficients in 1M NaNO3 was 1.4.106 L kg-1. All studied sorbents are suitable for deactivation of solutions containing EDTA. Cesium distribution coefficients were around 102-103 L kg-1 depending on EDTA concentration. Chemical stability of the sorbents was also studied. It was shown that 137Cs leaching rate from all sorbents meet the requirements for matrix materials.

  11. Immobilization of high-level defense waste in a slurry-fed electric glass melter

    International Nuclear Information System (INIS)

    Brouns, R.A.; Mellinger, G.B.; Nelson, T.A.; Oma, K.H.

    1980-11-01

    Scoping studies have been performed at the Pacific Northwest Laboratory related to the direct liquid-feeding of a generic high-level defense waste to a joule-heated ceramic melter. Tests beginning on the laboratory scale and progressing to full-scale operation are reported. Laboratory work identified the need for a reducing agent in the feed to help control the foaming tendencies of the waste glass. These tests also indicated that suspension agents were helpful in reducing the tendency of solids to settle out of the liquid feed. Testing was then moved to a larger pilot-scale melter (designed for approx. 2.5 kg/h) where verification of the flowsheet examined in the lab was accomplished. It was found that the reducing agent controlled foaming and did not result in the precipitation of metals. Pumping problems were encountered when slurries with higher than normal solids content were fed. A demonstration (designed for approx. 50 kg/h) in a full-scale melter was then made with the tested flowsheet; however, the amount of reducing agent had to be increased. In addition, it was found that feed control needed further development; however, steady-state operation was achieved giving encouraging results on process capacities. During steady-state operation, ruthenium losses to the offgas system averaged less than 0.16%, while cesium losses were somewhat higher, ranging from 0.91 to 24% and averaging 13%. Particulate decontamination factors from feed to offgas in the melter ranged from 5 x 10 2 to greater than 10 3 without any filtration or treatment. Approximately 1050 kg of glass was produced from 2900 L of waste at rates up to 40 kg/h

  12. Systems engineering management and implementation plan for Project W-464, immobilized high-level waste storage

    International Nuclear Information System (INIS)

    Wecks, M.D.

    1998-01-01

    The Systems Engineering Management and Implementation Plan (SEMIP) for TWRS Project W-46 describes the project implementation of the Tank Waste Remediation System Systems Engineering Management Plan. (TWRS SEMP), Rev. 1. The SEMIP outlines systems engineering (SE) products and processes to be used by the project for technical baseline development. A formal graded approach is used to determine the products necessary for requirements, design, and operational baseline completion. SE management processes are defined, and roles and responsibilities for management processes and major technical baseline elements are documented

  13. Systems engineering management and implementation plan for Project W-465, immobilized low-activity waste storage

    International Nuclear Information System (INIS)

    Kaspar, J.R.; Latray, D.A.

    1998-01-01

    The Systems Engineering Management and Implementation Plan (SEMIP) for TWRS Project W-465 describes the project implementation of the Tank Waste Remediation System Systems Engineering Management Plan (TWRS SEMP), Rev. 1. The SEMIP outlines systems engineering (SE) products and processes to be used by the project for technical baseline development. A formal graded approach is used to determine the products necessary for requirements, design, and operational baseline completion. SE management processes are defined, and roles and responsibilities for management processes and major technical baseline elements are documented

  14. Immobilization of high level nuclear reactor wastes in SYNROC: a current appraisal

    International Nuclear Information System (INIS)

    Oversby, V.M.; Ringwood, A.E.

    1981-01-01

    Results are presented for leach testing at 95 0 C and 200 0 C of SYNROC containing 9% and 20% simulated high level radioactive waste, synthetic hollandite and pervoskite samples, and natural zirconolite and pervoskite samples. Single phase synthetic minerals show much higher leach rates than natural mineral samples and polyphase SYNROC samples. Natural zirconolite samples with low radiation damage have leach rates at 200 0 C based on U which are identical to those measured on SYNROC samples. Natural zirconolites with very large accumulated α dose and radiation damage have leach rates at 200 0 C which are only 5 times higher than those of low dose samples

  15. Immobilization of strontium and cesium in intermediate-level liquid wastes by solidification in cements

    International Nuclear Information System (INIS)

    Rudolph, G.; Koester, R.

    1979-01-01

    An accelerated leach test at elevated temperature has been developed which gives intercomparable results within one day. It is very useful for product quality control at large throughputs. Using this test, it has been shown that cesium leachabilities from cement products containing a simulated waste typical of fuel reprocessing plants can be reduced by addition of a bentonite. Addition of barium silicate hydrate retards strontium leaching in these cements. Leach rates in tap water and in salt brine are lower than in distilled water and sodium chloride solution

  16. Immobilization of sodium and phosphorus-bearing PW-7a waste in SYNROC. Progress report

    International Nuclear Information System (INIS)

    Ringwood, A.E.

    1982-01-01

    The phosphorus, sodium and gadolinium-rich PW-7a waste can be successfully incorporated in SYNROC-C. However, a new accessory phase, a Ca,Na,Ba phosphate isostructural with Ca 5 Na 2 (PO 4 ) 4 apppears in the SYNROC mineralogy. There is no evidence for the partition of key radionuclides (e.g. Sr, REE and hence actinides) into this phosphate. Its poor resistance to groundwater dissolution, whilst hardly desirable, may therefore not have a serious effect on the leaching performance of SYNROC containing PW-7a. 9 tables

  17. The possibility use estimate of the concrete-polymers for immobilization of radioactive wastes

    International Nuclear Information System (INIS)

    Kapustina, I.B.; Starchenko, T.V.

    1994-01-01

    One of main ways of washability decrease of radionuclides is a reduction of cement stone porosity. With this purpose it is reduced water-cement attitude with 0.7 till 0.35, that, however, results in deterioration of cement stone fluidity, or is carried out impregnation of cement by monomers. For improvement of the cement block characteristics with included radioactive waste an opportunity of application of a new radiation way of manufacture of concrete-polymers is investigated. Essence of a way consists of impregnation concrete matrix by nontoxic and nonvolatile oligo-esters, polymerizing with formation of mesh polymers. In result of such processing is received compound material, having increased strength, radiation and chemical stability, high resistance to cold and durability. The introduction of radioactive waste simulators in an initial composition results in significant reduction of concrete strength, while the impregnation of concrete by oligo-ester with subsequent polymerization increases strength of concrete without simulators in 2-1.8 times and in 2.5-3 times with ones. Thus concrete-polymer can become a reliable protective barrier on a way of allocation radioactivity from the block. 2 tabs., 2 figs., 8 refs

  18. Leaching and mechanical properties of cabal glasses developed as matrices for immobilization high-level wastes

    International Nuclear Information System (INIS)

    Ezz-Eldin, F.M.

    2001-01-01

    This paper discusses the leaching behavior of simulated high-level-waste cabal glass (CaO-B 2 O 3 -Al 2 O 3 ) as a bulk specimen. During leach tests, the glass is immersed in either deionized water or in groundwater for up to 57 days at 70 deg. C. Based on the results, mechanisms observed with the leaching of the glass in deionized water or groundwater are discussed. Three factors, i.e., time of immersion, type of leaching solution and irradiation effect, are extensively studied. The corrosion was found to be linear with time in the limit of investigation (1-57 days) but with different rates depending on the type of solution and glass composition. Effects of γ-irradiation on the glass together with groundwater were found to decrease the glass durability. The evolution of the damage on mechanical and physical properties of the glass before and after leaching or irradiation was also discussed. The addition of waste oxide changes the properties of the glass matrix, so the influence of the guest oxides on the properties of host materials is also discussed

  19. Design and operation of small-scale glass melters for immobilizing radioactive waste

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Chismar, P.H.

    1980-01-01

    A small-scale (3-kg), joule-heated, continuous melter has been designed to study vitrification of Savannah River Plant radioactive waste. The first melter built has been in nonradioactive service for nearly three years. This melter had Inconel 690 electrodes and uses Monofrax K-3 for the contact refractory. Several problems seem in this melter have had an impact on the design of a full-scale system. Problems include uncontrolled electric currents passing through the throat, and formation of a slag layer at the bottom of the melter. The performance of a similar melter in a low-maintenance, radioactive environment is also described. Problems such as halide refluxing, and hot streaking, first observed in this melter, are also discussed

  20. Immobilization of antimony in waste-to-energy bottom ash by addition of calcium and iron containing additives.

    Science.gov (United States)

    Van Caneghem, Jo; Verbinnen, Bram; Cornelis, Geert; de Wijs, Joost; Mulder, Rob; Billen, Pieter; Vandecasteele, Carlo

    2016-08-01

    The leaching of Sb from waste-to-energy (WtE) bottom ash (BA) often exceeds the Dutch limit value of 0.32mgkg(-1) for recycling of BA in open construction applications. From the immobilization mechanisms described in the literature, it could be concluded that both Ca and Fe play an important role in the immobilization of Sb in WtE BA. Therefore, Ca and Fe containing compounds were added to the samples of the sand fraction of WtE BA, which in contrast to the granulate fraction is not recyclable to date, and the effect on the Sb leaching was studied by means of batch leaching tests. Results showed that addition of 0.5 and 2.5% CaO, 5% CaCl2, 2.5% Fe2(SO4)3 and 1% FeCl3 decreased the Sb leaching from 0.62±0.02mgkgDM(-1) to 0.20±0.02, 0.083±0.044, 0.25±0.01, 0.27±0.002 and 0.29±0.02mgkgDM(-1), respectively. Due to the increase in pH from 11.41 to 12.53 when 2.5% CaO was added, Pb and Zn leaching increased and exceeded the respective leaching limits. Addition of 5% CaCO3 had almost no effect on the Sb leaching, as evidenced by the resulting 0.53mgkgDM(-1) leaching concentration. This paper shows a complementary enhancement of the effect of Ca and Fe, by comparing the aforementioned Sb leaching results with those of WtE BA with combined addition of 2.5% CaO or 5% CaCl2 with 2.5% Fe2(SO4)3 or 1% FeCl3. These lab scale results suggest that formation of romeites with a high Ca content and formation of iron antimonate (tripuhyite) with a very low solubility are the main immobilization mechanisms of Sb in WtE BA. Besides the pure compounds and their mixtures, also addition of 10% of two Ca and Fe containing residues of the steel industry, hereafter referred to as R1 and R2, was effective in decreasing the Sb leaching from WtE BA below the Dutch limit value for reuse in open construction applications. To evaluate the long term effect of the additives, pilot plots of WtE BA with 10% of R1 and 5% and 10% of R2 were built and samples were submitted to leaching tests at

  1. Contamination immobilization by polymer coating of large-sized radioactive wastes for storage

    International Nuclear Information System (INIS)

    Tassigny, C. de; Signoret, C.

    1990-01-01

    The aim of the research work is the development of new techniques to fix contamination and the containment of radioactive waste by using polyurethane paints. The following main work steps have been carried out: . Determination of the polyurethane paint most suited for this application based on water diffusion tests and diffusion of radio-elements . Choice of the spraying technique with respect to conditions of an ionizing environment . Pilot application of the technique on a dismantling site. The polyurethane paint selected is a solventless bicomponent paint manufactured by Bayer. The spraying device chosen (Isotherm PMS 70) and fitted with a Gusmer-Getrasur spray gun is able to operate in an ionizing environment under satisfactory conditions (tested on two pilot work sites). The process generates no contaminated aerosols, but ensures perfect fixation of the ionizing particles. ANDRA has given its agreement for storage in one of its sites of 12 alpha contaminated cast iron slabs (total weight: 60 tons) on which one ton of polyurethane has been applied

  2. Leaching studies of heavy concrete material for nuclear fuel waste immobilization containers

    International Nuclear Information System (INIS)

    Onofrei, M.; Raine, D.; Brown, L.; Hooton, R.D.

    1989-08-01

    The leaching behaviour of a high-density concrete was studied as part of a program to evaluate its potential use as a container material for nuclear fuel waste under conditions of deep geologic disposal. Samples of concrete material were leached in deionized distilled water, Standard Canadian Shield Saline Solution (SCSSS), SCSSS plus 20% Na-bentonite, and SCSSS plus granite and 20% Na-bentonite under static conditions at 100 degrees celsius for periods up to 365 days. The results of these leaching experiments suggest that the stability of concrete depends on the possible internal structural changes due to hydration reactions of unhydrated components, leading to the formation of C-S-H gel plus portlandite (Ca(OH) 2 ). The factors controlling the concrete leaching process were the composition of the leachant and the concentration of elements in solution capable of forming precipitates on the concrete surface, e.g., silicon, Mg 2+ and Ca 2+ . The main effect observed during leaching was an increase in groundwater pH (from 7 to 9). However, the addition of Na-bentonite suppressed the normal tendency of the pH of the groundwater in contact with concrete to rise rapidly. It was shown that the solution concentration of elements released from the concrete, particularly potassium, increased in the presence of Na-bentonite

  3. Study of the surface crystallization and resistance to dissolution of niobium phosphate glasses for nuclear waste immobilization

    International Nuclear Information System (INIS)

    Vieira, Heveline

    2008-01-01

    The surface crystallization and the dissolution rate of three phosphate glass compositions containing different amounts of niobium oxide were studied. The glasses were named Nb30, Nb37, and Nb44 according to the nominal content of niobium oxide in the glass composition. The three compositions were evaluated keeping the P 2 O 5 /K 2 O ratio constant and varying the amount of Nb 2 O 5 . These glasses were produced by melting appropriate chemical compounds at 1500 deg C for 0.5 hour. The crystalline phases which were nucleated on the glass surface after heat treatment were determined by X-ray diffraction. The crystalline structures depend on the amount of niobium oxide in the glass composition. The crystal morphologies were observed by using an optical microscope, and their characteristics are specific for each kind of crystalline phase. The crystal growth rate and the surface nuclei density were determined for each glass composition, and they depend on each crystalline phase nucleated on the surface. From the differential thermal analysis curves it was determined that the Nb44 glass containing 46.5 mol por cent of niobium oxide is the most thermally stable against crystallization when compared to the Nb30 and Nb37 glasses. According to the activation energies determined for crystal growth on the surface of each glass type, the Nb44 glass can also be considered the most resistant one against crystallization. The dissolution rate for the Nb44 glass after 14 days immersed in an aqueous solution with pH equals to 7 at 90 deg C is the lowest (9.0 x 10 -7 g. cm -2 . day -1 ) when compared to the other two glass compositions. The dissolution rates in acidic and neutral solutions of all studied glasses meet the international standards for materials which can be used in the immobilization of nuclear wastes. (author)

  4. Plutonium Disposition by Immobilization

    International Nuclear Information System (INIS)

    Gould, T.; DiSabatino, A.; Mitchell, M.

    2000-01-01

    The ultimate goal of the Department of Energy (DOE) Immobilization Project is to develop, construct, and operate facilities that will immobilize between 17 to 50 tonnes (MT) of U.S. surplus weapons-usable plutonium materials in waste forms that meet the ''spent fuel'' standard and are acceptable for disposal in a geologic repository. Using the ceramic can-in-canister technology selected for immobilization, surplus plutonium materials will be chemically combined into ceramic forms which will be encapsulated within large canisters of high level waste (HLW) glass. Deployment of the immobilization capability should occur by 2008 and be completed within 10 years. In support of this goal, the DOE Office of Fissile Materials Disposition (MD) is conducting development and testing (D and T) activities at four DOE laboratories under the technical leadership of Lawrence Livermore National Laboratory (LLNL). The Savannah River Site has been selected as the site for the planned Plutonium Immobilization Plant (PIP). The D and T effort, now in its third year, will establish the technical bases for the design, construction, and operation of the U. S. capability to immobilize surplus plutonium in a suitable and cost-effective manner. Based on the D and T effort and on the development of a conceptual design of the PIP, automation is expected to play a key role in the design and operation of the Immobilization Plant. Automation and remote handling are needed to achieve required dose reduction and to enhance operational efficiency

  5. Single Phase Melt Processed Powellite (Ba,Ca) MoO{sub 4} For The Immobilization Of Mo-Rich Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, Kyle [Savannah River Site (SRS), Aiken, SC (United States); Marra, James [Savannah River Site (SRS), Aiken, SC (United States); Fox, Kevin [Savannah River Site (SRS), Aiken, SC (United States); Reppert, Jason [Savannah River Site (SRS), Aiken, SC (United States); Crum, Jarrod [Paci fic Northwest National Laboratory , Richland, WA (United States); Tang, Ming [Los Alamos National Laboratory , Los Alamos, NM (United States)

    2012-09-17

    Crystalline and glass composite materials are currently being investigated for the immobilization of combined High Level Waste (HLW) streams resulting from potential commercial fuel reprocessing scenarios. Several of these potential waste streams contain elevated levels of transition metal elements such as molybdenum (Mo). Molybdenum has limited solubility in typical silicate glasses used for nuclear waste immobilization. Under certain chemical and controlled cooling conditions, a powellite (Ba,Ca)MoO{sub 4} crystalline structure can be formed by reaction with alkaline earth elements. In this study, single phase BaMoO{sub 4} and CaMoO{sub 4} were formed from carbonate and oxide precursors demonstrating the viability of Mo incorporation into glass, crystalline or glass composite materials by a melt and crystallization process. X-ray diffraction, photoluminescence, and Raman spectroscopy indicated a long range ordered crystalline structure. In-situ electron irradiation studies indicated that both CaMoO{sub 4} and BaMoO{sub 4} powellite phases exhibit radiation stability up to 1000 years at anticipated doses with a crystalline to amorphous transition observed after 1 X 10{sup 13} Gy. Aqueous durability determined from product consistency tests (PCT) showed low normalized release rates for Ba, Ca, and Mo (<0.05 g/m{sup 2}).

  6. Waste Treatment and Immobilization Plant U. S. Department of Energy Office of River Protection Submerged Bed Scrubber Condensate Disposition Project - 13460

    Energy Technology Data Exchange (ETDEWEB)

    Yanochko, Ronald M. [Washington River Protection Solutions, P.O. Box 850, Richland, Washington 99352 (United States); Corcoran, Connie [AEM Consulting, LLC, 1201 Jadwin Avenue, Richland, Washington 99352 (United States)

    2013-07-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) will generate an off-gas treatment system secondary liquid waste stream [submerged bed scrubber (SBS) condensate], which is currently planned for recycle back to the WTP Low Activity Waste (LAW) melter. This SBS condensate waste stream is high in Tc-99, which is not efficiently captured in the vitrified glass matrix [1]. A pre-conceptual engineering study was prepared in fiscal year 2012 to evaluate alternate flow paths for melter off-gas secondary liquid waste generated by the WTP LAW facility [2]. This study evaluated alternatives for direct off-site disposal of this SBS without pre-treatment, which mitigates potential issues associated with recycling. This study [2] concluded that SBS direct disposal is a viable option to the WTP baseline. The results show: - Off-site transportation and disposal of the SBS condensate is achievable and cost effective. - Reduction of approximately 4,325 vitrified WTP Low Activity Waste canisters could be realized. - Positive WTP operational impacts; minimal WTP construction impacts are realized. - Reduction of mass flow from the LAW Facility to the Pretreatment Facility by 66%. - Improved Double Shell Tank (DST) space management is a benefit. (authors)

  7. Physicochemical characterization of the yeast cells and the waste lignocellulosic particles in the immobilization process for ethanol production

    DEFF Research Database (Denmark)

    Agudelo-Escobar, Lina María; Mussatto, Solange I.; Peñuela, Mariana

    2017-01-01

    Ethanol is one of the leading alternative fuels. Efforts have increased the development of technologies for producing ethanol efficiently and economically. The continuous fermentation using yeast cells immobilized in low‐cost materials is presented as an excellent alternative. We used four...... to confirm the hydrophobic or hydrophilic character and the free energies interaction was established. Images were obtained by scanning electron microscope, and determination of surface areas and volumes was performed by adsorption and desorption isotherms. It was established that cell surface properties...... are modified by the immobilization process to which they are subjected. It was evident that cell immobilization depended on the properties of the carrier, as well as cell surface properties. Thus, in order to improve the process of cell immobilization, it is essential to understand the type of carrier‐cell...

  8. RHENIUM SOLUBILITY IN BOROSILICATE NUCLEAR WASTE GLASS IMPLICATIONS FOR THE PROCESSING AND IMMOBILIZATION OF TECHNETIUM-99 (AND SUPPORTING INFORMATION WITH GRAPHICAL ABSTRACT)

    Energy Technology Data Exchange (ETDEWEB)

    AA KRUGER; A GOEL; CP RODRIGUEZ; JS MCCLOY; MJ SCHWEIGER; WW LUKENS; JR, BJ RILEY; D KIM; M LIEZERS; P HRMA

    2012-08-13

    The immobilization of 99Tc in a suitable host matrix has proved a challenging task for researchers in the nuclear waste community around the world. At the Hanford site in Washington State in the U.S., the total amount of 99Tc in low-activity waste (LAW) is {approx} 1,300 kg and the current strategy is to immobilize the 99Tc in borosilicate glass with vitrification. In this context, the present article reports on the solubility and retention of rhenium, a nonradioactive surrogate for 99Tc, in a LAW sodium borosilicate glass. Due to the radioactive nature of technetium, rhenium was chosen as a simulant because of previously established similarities in ionic radii and other chemical aspects. The glasses containing target Re concentrations varying from 0 to10,000 ppm by mass were synthesized in vacuum-sealed quartz ampoules to minimize the loss of Re by volatilization during melting at 1000 DC. The rhenium was found to be present predominantly as Re7 + in all the glasses as observed by X-ray absorption near-edge structure (XANES). The solubility of Re in borosilicate glasses was determined to be {approx}3,000 ppm (by mass) using inductively coupled plasma-optical emission spectroscopy (ICP-OES). At higher rhenium concentrations, some additional material was retained in the glasses in the form of alkali perrhenate crystalline inclusions detected by X-ray diffraction (XRD) and laser ablation-ICP mass spectrometry (LA-ICP-MS). Assuming justifiably substantial similarities between Re7 + and Tc 7+ behavior in this glass system, these results implied that the processing and immobilization of 99Tc from radioactive wastes should not be limited by the solubility of 99Tc in borosilicate LAW glasses.

  9. Immobilization of acid digestion residue

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.; Allen, C.R.

    1983-01-01

    Acid digestion treatment of nuclear waste is similar to incineration processes and results in the bulk of the waste being reduced in volume and weight to some residual solids termed residue. The residue is composed of various dispersible solid materials and typically contains the resultant radioactivity from the waste. This report describes the immobilization of the residue in portland cement, borosilicate glass, and some other waste forms. Diagrams showing the cement and glass virtification parameters are included in the report as well as process steps and candidate waste product forms. Cement immobilization is simplest and probably least expensive; glass vitrification exhibits the best overall volume reduction ratio

  10. Immobilization of lead in a Korean military shooting range soil using eggshell waste: An integrated mechanistic approach

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, Mahtab [Department of Biological Environment, Kangwon National University, Chuncheon 200-701 (Korea, Republic of); Hashimoto, Yohey [Department of Bioresource Science, Mie University, 1577 Kurima-machiya, Mie 514-8507 (Japan); Moon, Deok Hyun [Department of Environmental Engineering, Chosun University, Gwangju 501-759 (Korea, Republic of); Lee, Sang Soo, E-mail: sslee97@kangwon.ac.kr [Department of Biological Environment, Kangwon National University, Chuncheon 200-701 (Korea, Republic of); Ok, Yong Sik, E-mail: soilok@kangwon.ac.kr [Department of Biological Environment, Kangwon National University, Chuncheon 200-701 (Korea, Republic of)

    2012-03-30

    Highlights: Black-Right-Pointing-Pointer Eggshell and calcined eggshell immobilized Pb in the shooting range soil. Black-Right-Pointing-Pointer Calcined eggshell was more effective on Pb immobilization compared to eggshell. Black-Right-Pointing-Pointer Exchangeable Pb fractions were transformed to carbonate bound fractions. Black-Right-Pointing-Pointer Calcined eggshell stabilized Pb by enwrapping into calcium silicate hydrate. Black-Right-Pointing-Pointer Soil Pb toxicity can be reduced by applying eggshell and calcined eggshell. - Abstract: This study evaluated the effectiveness of eggshell and calcined eggshell on lead (Pb) immobilization in a shooting range soil. Destructive and non-destructive analytical techniques were employed to determine the mechanism of Pb immobilization. The 5% additions of eggshell and calcined eggshell significantly decreased the TCLP-Pb concentration by 68.8% due mainly to increasing soil pH. Eggshell and calcined-eggshell amendments decreased the exchangeable Pb fraction to {approx}1% of the total Pb in the soil, while the carbonate-associated Pb fraction was increased to 40.0-47.1% at >15% application rates. The thermodynamic modeling on Pb speciation in the soil solution predicted the precipitation of Pb-hydroxide [Pb(OH){sub 2}] in soils amended with eggshell and calcined eggshell. The SEM-EDS, XAFS and elemental dot mapping revealed that Pb in soil amended with calcined eggshell was associated with Si and Ca, and may be immobilized by entrapping into calcium-silicate-hydrate. Comparatively, in the soil amended with eggshell, Pb was immobilized via formation of Pb-hydroxide or lanarkite [Pb{sub 2}O(SO{sub 4})]. Applications of amendments increased activities of alkaline phosphatase up to 3.7 times greater than in the control soil. The use of eggshell amendments may have potential as an integrated remediation strategy that enables Pb immobilization and soil biological restoration in shooting range soils.

  11. Immobilization of lead in a Korean military shooting range soil using eggshell waste: An integrated mechanistic approach

    International Nuclear Information System (INIS)

    Ahmad, Mahtab; Hashimoto, Yohey; Moon, Deok Hyun; Lee, Sang Soo; Ok, Yong Sik

    2012-01-01

    Highlights: ► Eggshell and calcined eggshell immobilized Pb in the shooting range soil. ► Calcined eggshell was more effective on Pb immobilization compared to eggshell. ► Exchangeable Pb fractions were transformed to carbonate bound fractions. ► Calcined eggshell stabilized Pb by enwrapping into calcium silicate hydrate. ► Soil Pb toxicity can be reduced by applying eggshell and calcined eggshell. - Abstract: This study evaluated the effectiveness of eggshell and calcined eggshell on lead (Pb) immobilization in a shooting range soil. Destructive and non-destructive analytical techniques were employed to determine the mechanism of Pb immobilization. The 5% additions of eggshell and calcined eggshell significantly decreased the TCLP-Pb concentration by 68.8% due mainly to increasing soil pH. Eggshell and calcined-eggshell amendments decreased the exchangeable Pb fraction to ∼1% of the total Pb in the soil, while the carbonate-associated Pb fraction was increased to 40.0–47.1% at >15% application rates. The thermodynamic modeling on Pb speciation in the soil solution predicted the precipitation of Pb-hydroxide [Pb(OH) 2 ] in soils amended with eggshell and calcined eggshell. The SEM-EDS, XAFS and elemental dot mapping revealed that Pb in soil amended with calcined eggshell was associated with Si and Ca, and may be immobilized by entrapping into calcium-silicate-hydrate. Comparatively, in the soil amended with eggshell, Pb was immobilized via formation of Pb-hydroxide or lanarkite [Pb 2 O(SO 4 )]. Applications of amendments increased activities of alkaline phosphatase up to 3.7 times greater than in the control soil. The use of eggshell amendments may have potential as an integrated remediation strategy that enables Pb immobilization and soil biological restoration in shooting range soils.

  12. A review of methods for immobilizing iodine-129 arising from a nuclear fuel recycle plant, with emphasis on waste-form chemistry

    International Nuclear Information System (INIS)

    Taylor, P.

    1990-07-01

    Possible methods for the separation and immobilization of iodine (mainly iodine-129) in a fuel recycle plant are reviewed, with special emphasis placed on the evaluation of waste forms. A distinction is drawn between waste forms selected by thermodynamic (solubility) or kinetic (dissolution rate) considerations. The most promising solubility-limited waste forms appear to be AgI (or AgI + AgCl) and a combination of Bi 2 O 3 and Bi 5 O 7 I. These materials use relatively scarce metals, Ag and Bi. They also have substantial chemical limitations, such as susceptibility to reductive dissolution and anion-displacement reactions; this calls for special care in the choice of a disposal site. All other organic iodides and iodates considered here and elsewhere appear to be still more limited in this respect. The most promising kinetically limited candidate waste form appears to be iodide-sodalite, but further information is needed on both the fabrication and leaching behaviour of this material. The possibility of disposal in a more soluble but isotopically dilute waste form, employing abundant raw materials, also warrants further consideration

  13. Immobilization of lead in a Korean military shooting range soil using eggshell waste: an integrated mechanistic approach.

    Science.gov (United States)

    Ahmad, Mahtab; Hashimoto, Yohey; Moon, Deok Hyun; Lee, Sang Soo; Ok, Yong Sik

    2012-03-30

    This study evaluated the effectiveness of eggshell and calcined eggshell on lead (Pb) immobilization in a shooting range soil. Destructive and non-destructive analytical techniques were employed to determine the mechanism of Pb immobilization. The 5% additions of eggshell and calcined eggshell significantly decreased the TCLP-Pb concentration by 68.8% due mainly to increasing soil pH. Eggshell and calcined-eggshell amendments decreased the exchangeable Pb fraction to ≈ 1% of the total Pb in the soil, while the carbonate-associated Pb fraction was increased to 40.0-47.1% at >15% application rates. The thermodynamic modeling on Pb speciation in the soil solution predicted the precipitation of Pb-hydroxide [Pb(OH)(2)] in soils amended with eggshell and calcined eggshell. The SEM-EDS, XAFS and elemental dot mapping revealed that Pb in soil amended with calcined eggshell was associated with Si and Ca, and may be immobilized by entrapping into calcium-silicate-hydrate. Comparatively, in the soil amended with eggshell, Pb was immobilized via formation of Pb-hydroxide or lanarkite [Pb(2)O(SO(4))]. Applications of amendments increased activities of alkaline phosphatase up to 3.7 times greater than in the control soil. The use of eggshell amendments may have potential as an integrated remediation strategy that enables Pb immobilization and soil biological restoration in shooting range soils. Copyright © 2012 Elsevier B.V. All rights reserved.

  14. Immobilization of LiCl-Li 2 O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

    2018-04-01

    In this study, salt occlusion and hydrothermal processes were used to make chlorosodalite through reaction with a high-LiCl salt simulating a waste stream following pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and aide in densification. Hydrothermal processes included reaction of the salt simulant in an acid digestion vessel with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

  15. Immobilization of LiCl-Li2O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

    Science.gov (United States)

    Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

    2018-04-01

    In this study, hydrothermal and salt-occlusion processes were used to make chlorosodalite through reactions with a high-LiCl salt simulating a waste stream generated from pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and to aid in densification. Hydrothermal processes included reaction of the salt simulant in an autoclave with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

  16. Immobilization of Iron Nanoparticles on Multi Substrates and Its Reduction Removal of Chromium (VI) from Waste Streams

    Science.gov (United States)

    This article describes the in-situ synthesis and immobilization of iron nanoparticles on several substrates at room temperature using NaBH4 as a reducing agent and ascorbic acid as capping agent. The method is very effective in protecting iron nanoparticles from air oxidation for...

  17. Effect of organic loading rate on dark fermentative hydrogen production in the continuous stirred tank reactor and continuous mixed immobilized sludge reactor from waste pastry hydrolysate.

    Science.gov (United States)

    Han, Wei; Hu, Yunyi; Li, Shiyi; Nie, Qiulin; Zhao, Hongting; Tang, Junhong

    2016-12-01

    Waste pastry (6%, w/v) was hydrolyzed by the produced glucoamylase and protease to obtain the glucose (19.8g/L) and free amino nitrogen (179mg/L) solution. Then, the effect of organic loading rate (OLR) (8-40kgCOD/(m 3 d)) on dark fermentative hydrogen production in the continuous stirred tank reactor (CSTR) and continuous mixed immobilized sludge reactor (CMISR) from waste pastry hydrolysate was investigated and compared. The maximum hydrogen production rate of CSTR (277.76mL/(hL)) and CMISR (320.2mL/(hL)) were achieved at OLR of 24kgCOD/(m 3 d) and 32kgCOD/(m 3 d), respectively. Carbon recovery ranged from 75.2-84.1% in the CSTR and CMISR with the balance assumed to be converted to biomass. One gram waste pastry could produce 0.33g (1.83mmol) glucose which could be further converted to 79.24mL (3.54mmol) hydrogen in the CMISR or 91.66mL (4.09mmol) hydrogen in the CSTR. This is the first study which reports dark fermentative hydrogen production from waste pastry. Copyright © 2016 Elsevier Ltd. All rights reserved.

  18. Weldon Spring, Missouri, Raffinate Pits 1, 2, 3, and 4: Preliminary grout development screening studies for in situ waste immobilization

    International Nuclear Information System (INIS)

    McDaniel, E.W.; Gilliam, T.M.; Dole, L.R.; West, G.A.

    1987-04-01

    Results of Oak Ridge National Laboratory's initial support program to develop a preliminary grout formula to solidify in situ the Weldon Spring waste are presented. The screening study developed preliminary formulas based on a simulated composite waste and then tested the formulas on actual waste samples. Future data needs are also discussed. 1 ref., 6 figs., 9 tabs

  19. Immobilization of bacteria selected for the removal of toxic waste trapped in hydrogels obtained by ionizing radiation

    International Nuclear Information System (INIS)

    Fernandez Degiorgi, Cristina H.C.; Pizarro, Ramon A.; Fernandez, Ruben O.; Carenza, M.; Lora, S.; Smolko, Eduardo E.

    1999-01-01

    Bacterial strains capable of growing in the presence of heavy metals were selected from soil and water from the Rio de la Plata coasts in Argentina and cultured in the hydrophilic membranes with the aim of bioremediation of the standard contaminated solutions. Bacterial cells were immobilized in polymeric matrices prepared by gamma irradiation of 2-hydroxyethyl methacrylate and 2-hydroxyethyl acrylate at -78 C degrees in the presence of water and glycerol and examined as carriers for cells immobilization in metal decontamination experiments. The results obtained indicate that removal from free bacteria was more efficient for Pb(II) and Cd(II) than for Cr(III) and Cu(II). Bacterial adhesion to hydrogels evaluated by scanning microscopic electronic was satisfactory leading the suitable biomass mechanical firmness. (author)

  20. A THERMAL MODEL OF THE IMMOBILIZATION OF LOW-LEVEL RADIOACTIVE WASTE AS GROUT IN CONCRETE VAULTS

    Energy Technology Data Exchange (ETDEWEB)

    Shadday, M

    2008-10-27

    Salt solution will be mixed with cement and flyash/slag to form a grout which will be immobilized in above ground concrete vaults. The curing process is exothermic, and a transient thermal model of the pouring and curing process is herein described. A peak temperature limit of 85 C for the curing grout restricts the rate at which it can be poured into a vault. The model is used to optimize the pouring.

  1. Multibarrier effectiveness as the expedient measure for selecting the appropriate stabilization and immobilization procedure for the various waste categories

    International Nuclear Information System (INIS)

    Merz, E.P.

    1998-01-01

    The management of radioactive wastes has become a major concern particularly with regard to the release of radioactive material to the environment and possible risks of contamination. The development of rational and acceptable options for radioactive waste disposal requires a clear understanding of radiators protection objectives and their application in planning, regulation and licensing. Considerable progress has been made over the past three decades within many countries utilising nuclear power to develop strategies for the management of nuclear wastes. All wastes should be managed in such a way that high standards of conditioning are maintained and that potential hazards originating from their disposal are reduced to levels that are as low as reasonable and well below admissible levels. However, deficiencies are evident in some areas of nuclear weapon fabrication. The nuclear fuel cycle is associated in the military weapon fabrication sector as well as in the civilian energy production field with two rather similar types of risk: 1. the risk due to the operation of the nuclear reactors and the appertaining fuel facilities, and 2. the risk contribution originating from the generation of radioactive wastes. The difference between these two categories of risk is that the first one has only a short time factor associated with it, since the lifetime of the plants is relatively short and drops to zero after plant shutdown. The second category is, more or less, a permanent kind of risk which will be inherited by future generations. Actual health effects of waste on people and populations, particularly over long periods of time, are not necessarily related to the level of radioactivity. If intensely radioactive waste is effectively isolated, then the radiation dose it causes can be much less than that accumulating from widely-dispersed but low-activity waste, particularly if this includes long-lived radioisotopes. By far the most important producers of nuclear wastes

  2. Aerosol Formation from High-Pressure Sprays for Supporting the Safety Analysis for the Hanford Waste Treatment and Immobilization Plant - 13183

    Energy Technology Data Exchange (ETDEWEB)

    Gauglitz, P.A.; Mahoney, L.A.; Schonewill, P.P.; Bontha, J.R.; Blanchard, J.; Kurath, D.E.; Daniel, R.C.; Song, C. [Pacific Northwest National Laboratory, PO Box 999, Richland WA 99352 (United States)

    2013-07-01

    The Waste Treatment and Immobilization Plant (WTP) at Hanford is being designed and built to pretreat and vitrify waste currently stored in underground tanks at Hanford. One of the postulated events in the hazard analysis for the WTP is a breach in process piping that produces a pressurized spray with small droplets that can be transported into ventilation systems. Literature correlations are currently used for estimating the generation rate and size distribution of aerosol droplets in postulated releases. These correlations, however, are based on results obtained from small engineered nozzles using Newtonian liquids that do not contain slurry particles and thus do not represent the fluids and breaches in the WTP. A test program was developed to measure the generation rate, and the release fraction which is the ratio of generation rate to spray flow rate, of droplets suspended in a test chamber and droplet size distribution from prototypic sprays. A novel test method was developed to allow measurement of sprays from small to large breaches and also includes the effect of aerosol generation from splatter when the spray impacts on walls. Results show that the release fraction decreases with increasing orifice area, though with a weaker dependence on orifice area than the currently-used correlation. A comparison of water sprays to slurry sprays with 8 to 20 wt% gibbsite or boehmite particles shows that the presence of slurry particles depresses the release fraction compared to water for droplets above 10 μm and increases the release fraction below this droplet size. (authors)

  3. Crossflow Ultra-filter Module Draining and Flush Testing for the Hanford Tank Waste Treatment and Immobilization Plant - Lessons Learned in De-clogging Crossflow Filters

    International Nuclear Information System (INIS)

    Townson, P.S.; Brackenbury, P.J.

    2009-01-01

    This paper describes test work conducted in order to study crossflow ultra-filter module draining and flushing for the Hanford Tank Waste Treatment and Immobilization Plant. The objective of the testing was to demonstrate that the current design, with a flush tank at elevation 29.9 m (98'-00'') has enough pressure head to drain (to a minimum elevation ∼1.5 m [∼5'-00'']) and clean out the ultra-filter tube side. Without demonstrating this, a potential failure of the flush system could cause immovable solids to plug the tubular membranes of the filters causing serious adverse impacts to plant availability and/or throughput, and could permit deleterious flammable gas accumulations. In conjunction with the water flush, the plant also utilizes air purging to prevent build up of flammable gases. Two filter configurations were investigated, one being the baseline horizontal layout and one being an alternative vertical layout. The slurry used in the tests was a non radioactive simulant (kaolin-bentonite clay), and it mimicked the rheological properties of the real waste slurry. The filter modules were full scale items, being 2.44 m (8') in length and containing 241 by 1.3 cm (1/2'') id sintered stainless steel filter tubes. (authors)

  4. Microorganism immobilization

    Science.gov (United States)

    Compere, Alicia L.; Griffith, William L.

    1981-01-01

    Live metabolically active microorganisms are immobilized on a solid support by contacting particles of aggregate material with a water dispersible polyelectrolyte such as gelatin, crosslinking the polyelectrolyte by reacting it with a crosslinking agent such as glutaraldehyde to provide a crosslinked coating on the particles of aggregate material, contacting the coated particles with live microorganisms and incubating the microorganisms in contact with the crosslinked coating to provide a coating of metabolically active microorganisms. The immobilized microorganisms have continued growth and reproduction functions.

  5. Anaerobic treatment of palm oil mill effluent in batch reactor with digested biodiesel waste as starter and natural zeolite for microbial immobilization

    Science.gov (United States)

    Setyowati, Paulina Adina Hari; Halim, Lenny; Mellyanawaty, Melly; Sudibyo, Hanifrahmawan; Budhijanto, Wiratni

    2017-05-01

    Palm oil mill effluent (POME) is the wastewater discharged from sludge separation, sterilization, and clarification process of palm oil industries. Each ton of palm oil produces about half ton of high organic load wastewater. Up to now, POME treatment is done in lagoon, leaving major problems in land requirement and greenhouse gasses release. The increasing of palm oil production provokes the urgency of appropriate technology application in treating POME to prevent the greenhouse gasses emission while exploit POME as renewable energy source. The purposes of this study were firstly to test the effectiveness of using the digested biodiesel waste as the inoculum and secondly to evaluate the effectiveness of natural zeolite addition in minimizing the inhibitory effect in digesting POME. It was expected that the oil-degrading bacteria in the inoculum would shorten the adaptation period in digesting POME. Furthermore, the consortium formation of anaerobic bacteria accelerated by natural zeolite powder addition would increase the microbial resistance to the inhibitors contained in the POME. The batch digesters, containing 0 (control); 17; 38; and 63 g natural zeolite/g sCOD substrate were observed for 43 days. The result showed that zeolite addition did not give significant effect on sCOD reduction (97.3-98.6% of initial sCOD). Moreover, addition of immobilization media up to 17 g natural zeolite/g stimulated the acidification and biogas production up to 10% higher than control. The purity of methane produced with various amount of immobilization media did not differ for each variation, i.e. 50-54% v/v methane. The increasing amount of natural zeolite up to 63 g/g sCOD did not significantly enhance biogas product rate nor methane content.

  6. Preliminary evaluation of the immobilization of simulated evaporator concentrate waste in low density polyethylene by extrusion process

    International Nuclear Information System (INIS)

    Cota, Stela; Oliveira, Tania Valeria S. de; Senne Junior, Murillo; Pacheco, Graziella

    2007-01-01

    Simulated evaporator concentrate was prepared by pre-treating sodium borate with calcium hydroxide to produce an insoluble borate salt. The resultant solid waste was blended by extrusion with virgin low density polyethylene (LDPE) in the proportion of 30 wt%. Samples were prepared to evaluate homogeneity, mechanical strength and leaching behavior. The homogeneity of each sample individually and in consecutive samples was indirectly estimated by sectioning each sample in four pieces and submitting each piece to density determination (ASTM standard D-792). Mechanical strength was evaluated through determination of compressive strength (ASTM standard D-695), and the results were compared to the value for the pure polymer and with the limit established by CNEN standard NN-6.09 for cement waste products. Samples were also tested for leaching by accelerated leaching test (ASTM standard C1308). Results showed a good homogeneity. Standard deviations of the density measurements were less than 1% for a single sample and less than 6% considering 3 samples. Polymer compressive strength at yield point and at 5% and 10% strain have increased after the mixture with the simulated waste, indicating an increase on the material strength. Estimated compressive strength was above CNEN standard limit for cement waste products if 5% strain could be considered a reasonable limit to assure structural integrity of the material. Cumulated leaching fraction after 11 days of accelerated leaching test was found to be below 10%, and diffusion coefficient was estimated as 9.06 x 10 -10 cm 2 /s, with deviation of 8.3%. (author)

  7. Preliminary evaluation of the immobilization of simulated evaporator concentrate waste in low density polyethylene by extrusion process

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Stela; Oliveira, Tania Valeria S. de; Senne Junior, Murillo; Pacheco, Graziella [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mail: sdsc@cdtn.br

    2007-07-01

    Simulated evaporator concentrate was prepared by pre-treating sodium borate with calcium hydroxide to produce an insoluble borate salt. The resultant solid waste was blended by extrusion with virgin low density polyethylene (LDPE) in the proportion of 30 wt%. Samples were prepared to evaluate homogeneity, mechanical strength and leaching behavior. The homogeneity of each sample individually and in consecutive samples was indirectly estimated by sectioning each sample in four pieces and submitting each piece to density determination (ASTM standard D-792). Mechanical strength was evaluated through determination of compressive strength (ASTM standard D-695), and the results were compared to the value for the pure polymer and with the limit established by CNEN standard NN-6.09 for cement waste products. Samples were also tested for leaching by accelerated leaching test (ASTM standard C1308). Results showed a good homogeneity. Standard deviations of the density measurements were less than 1% for a single sample and less than 6% considering 3 samples. Polymer compressive strength at yield point and at 5% and 10% strain have increased after the mixture with the simulated waste, indicating an increase on the material strength. Estimated compressive strength was above CNEN standard limit for cement waste products if 5% strain could be considered a reasonable limit to assure structural integrity of the material. Cumulated leaching fraction after 11 days of accelerated leaching test was found to be below 10%, and diffusion coefficient was estimated as 9.06 x 10{sup -10} cm{sup 2}/s, with deviation of 8.3%. (author)

  8. Bovine pancreatic trypsin inhibitor immobilized onto sepharose as a new strategy to purify a thermostable alkaline peptidase from cobia (Rachycentron canadum) processing waste.

    Science.gov (United States)

    França, Renata Cristina da Penha; Assis, Caio Rodrigo Dias; Santos, Juliana Ferreira; Torquato, Ricardo José Soares; Tanaka, Aparecida Sadae; Hirata, Izaura Yoshico; Assis, Diego Magno; Juliano, Maria Aparecida; Cavalli, Ronaldo Olivera; Carvalho, Luiz Bezerra de; Bezerra, Ranilson Souza

    2016-10-15

    A thermostable alkaline peptidase was purified from the processing waste of cobia (Rachycentron canadum) using bovine pancreatic trypsin inhibitor (BPTI) immobilized onto Sepharose. The purified enzyme had an apparent molecular mass of 24kDa by both sodium dodecyl sulfate polyacrylamide gel electrophoresis (SDS-PAGE) and mass spectrometry. Its optimal temperature and pH were 50°C and 8.5, respectively. The enzyme was thermostable until 55°C and its activity was strongly inhibited by the classic trypsin inhibitors N-ρ-tosyl-l-lysine chloromethyl ketone (TLCK) and benzamidine. BPTI column allowed at least 15 assays without loss of efficacy. The purified enzyme was identified as a trypsin and the N-terminal amino acid sequence of this trypsin was IVGGYECTPHSQAHQVSLNSGYHFC, which was highly homologous to trypsin from cold water fish species. Using Nα-benzoyl-dl-arginine ρ-nitroanilide hydrochloride (BApNA) as substrate, the apparent km value of the purified trypsin was 0.38mM, kcat value was 3.14s(-1), and kcat/km was 8.26s(-1)mM(-1). The catalytic proficiency of the purified enzyme was 2.75×10(12)M(-1) showing higher affinity for the substrate at the transition state than other fish trypsin. The activation energy (AE) of the BApNA hydrolysis catalyzed by this enzyme was estimated to be 11.93kcalmol(-1) while the resulting rate enhancement of this reaction was found to be approximately in a range from 10(9) to 10(10)-fold evidencing its efficiency in comparison to other trypsin. This new purification strategy showed to be appropriate to obtain an alkaline peptidase from cobia processing waste with high purification degree. According with N-terminal homology and kinetic parameters, R. canadum trypsin may gathers desirable properties of psychrophilic and thermostable enzymes. Copyright © 2016 Elsevier B.V. All rights reserved.

  9. Synthetic murataite-3C, a complex form for long-term immobilization of nuclear waste. Crystal structure and its comparison with natural analogues

    Energy Technology Data Exchange (ETDEWEB)

    Pakhomova, Anna S.; Krivovichev, Sergey V. [St. Petersburg State Univ. (Russian Federation). Dept. of Crystallography; Yudintsev, Sergey V. [Institute of Geology of Ore Deposits, Petrography, Mineralogy and Geochemistry, St. Petersburg (Russian Federation); Stefanovsky, Sergey V. [MosNPO Radon, Moscow (Russian Federation)

    2013-03-01

    The structure of synthetic murataite-3C intended for long-term immobilization of high-level radioactive waste has been solved using crystals prepared by melting in an electric furnace at 1500 C. The material is cubic, F- anti 43m, a = 14.676(15) A, V = 3161.31(57) A{sup 3}. The structure is based upon a three-dimensional framework consisting of {alpha}-Keggin [Al{sup [4]}Ti{sub 12}{sup [6]}O{sub 40}] clusters linked by sharing the O5 atoms. The Keggin-cluster-framework interpenetrates with the metal-oxide substructure that can be considered as a derivative of the fluorite structure. The crystal chemical formula of synthetic murataite-3C derived from the obtained structure model can be written as {sup [8]}Ca{sub 6}{sup [8]}Ca{sub 4}{sup [6]}Ti{sub 12}{sup [5]}Ti{sub 4}{sup [4]}AlO{sub 42}. Its comparison with the natural murataite shows that the synthetic material has a noticeably less number of vacancies in the cation substructure and contains five instead of four symmetrically independent cation positions. The presence of the additional site essentially increases the capacity of synthetic murataite with respect to large heavy cations such as actinides, rare earth and alkaline earth metals in comparison with the material of natural origin. (orig.)

  10. A thermal model of the immobilization of low-level radioactive waste as grout in concrete vaults

    International Nuclear Information System (INIS)

    Shadday, Martin A.

    2009-01-01

    Salt solution, from radioactive waste generated by the production of plutonium and tritium in nuclear reactors at the Savannah River Site, will be mixed with cement and flyash/slag to form a grout which will be poured into above ground concrete vaults. The curing process is exothermic, and a transient thermal model of the pouring and curing process is herein described. A peak temperature limit of 85 o C for the curing grout restricts the rate at which it can be poured into a vault. The model is used to optimize the pouring.

  11. IR and Raman spectroscopy of sodium aluminophosphate vitreous materials for immobilization of high level wastes from nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Stefanovskij, S.V.; Myasoedov, B.F.; Remizov, M.B.; Belanova, E.A.

    2014-01-01

    The structure of sodium aluminophosphate vitreous materials containing constituents of high level wastes (cesium, magnesium, copper and molybdenum oxides) from uranium-graphite reactor has been studied by IR and Raman spectroscopy techniques coupled with X-ray diffraction. The structural network has been shown to be composed of short phosphorus-oxygen chains with embedded aluminum-oxygen tetrahedra. In the magnesium-bearing samples the cross-linking with Mg 2+ ions is possible. The effect of other oxides (Cs 2 O, MoO 3 , CuO) on the glass structure is negligible for the occuring amounts. The glasses have been devitrified partly at the quenching and much stronger at the annealing. It is reflected in splitting of the vibration bands of the bonds in structural units of the anionic motif of the vitreous materials. (authors)

  12. Effect of composition and temperature on viscosity and electrical conductivity of borosilicate glasses for Hanford nuclear waste immobilization

    International Nuclear Information System (INIS)

    Hrma, P.; Piepel, G.F.; Smith, D.E.; Redgate, P.E.; Schweiger, M.J.

    1993-04-01

    Viscosity and electrical conductivity of 79 simulated borosilicate glasses in the expected range of compositions to be produced in the Hanford Waste Vitrification Plant were measured within the temperature span from 950 to 1250 degree C. The nine major oxide components were SiO 2 , B 2 O 3 , Li 2 O, Na 2 O, CaO, MgO, Fe 2 O 3 , Al 2 O 3 , and ZrO 2 . The test compositions were generated statistically. The data were fitted by Fulcher and Arrhenius equations with temperature coefficients being multilinear functions of the mass fractions of the oxide components. Mixture models were also developed for the natural logarithm of viscosity and that of electrical conductivity at 1150 degree C. Least squares regression was used to obtain component coefficients for all the models

  13. Technetium Immobilization Forms Literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H.; Cantrell, Kirk J.; Serne, R. Jeffrey; Qafoku, Nikolla

    2014-05-01

    Of the many radionuclides and contaminants in the tank wastes stored at the Hanford site, technetium-99 (99Tc) is one of the most challenging to effectively immobilize in a waste form for ultimate disposal. Within the Hanford Tank Waste Treatment and Immobilization Plant (WTP), the Tc will partition between both the high-level waste (HLW) and low-activity waste (LAW) fractions of the tank waste. The HLW fraction will be converted to a glass waste form in the HLW vitrification facility and the LAW fraction will be converted to another glass waste form in the LAW vitrification facility. In both vitrification facilities, the Tc is incorporated into the glass waste form but a significant fraction of the Tc volatilizes at the high glass-melting temperatures and is captured in the off-gas treatment systems at both facilities. The aqueous off-gas condensate solution containing the volatilized Tc is recycled and is added to the LAW glass melter feed. This recycle process is effective in increasing the loading of Tc in the LAW glass but it also disproportionally increases the sulfur and halides in the LAW melter feed which increases both the amount of LAW glass and either the duration of the LAW vitrification mission or the required supplemental LAW treatment capacity.

  14. FTIR and Mössbauer spectroscopic study of sodium–aluminum–iron phosphate glassy materials for high level waste immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovsky, S.V., E-mail: serge.stefanovsky@yandex.ru [Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences, Laboratory of Radioecology and Radiation Problems, Moscow (Russian Federation); Stefanovsky, O.I. [Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences, Laboratory of Radioecology and Radiation Problems, Moscow (Russian Federation); Remizov, M.B.; Belanova, E.A.; Kozlov, P.V. [FSUE PA Mayak, Central Plant Laboratory, Ozersk, Chelyabinsk Reg. (Russian Federation); Glazkova, Ya.S.; Sobolev, A.V.; Presniakov, I.A. [Lomonosov Moscow State University, Department of Radiochemistry (Russian Federation); Kalmykov, S.N. [Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences, Laboratory of Radioecology and Radiation Problems, Moscow (Russian Federation); Lomonosov Moscow State University, Department of Radiochemistry (Russian Federation); Vernadsky Institute of Geochemistry and Analytical Chemistry of the Russian Academy of Sciences, Laboratory of Radiochemistry, Moscow (Russian Federation); Myasoedov, B.F. [Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences, Laboratory of Radioecology and Radiation Problems, Moscow (Russian Federation); Vernadsky Institute of Geochemistry and Analytical Chemistry of the Russian Academy of Sciences, Laboratory of Radiochemistry, Moscow (Russian Federation)

    2015-11-15

    Complex sodium-aluminum-iron phosphate glassy materials with various Al{sub 2}O{sub 3} to Fe{sub 2}O{sub 3} ratio containing high level waste (HLW) surrogate were characterized by X-ray diffraction and scanning electron microscopy and studied in details by Fourier transform infrared (FTIR) spectroscopy. The samples with high Al{sub 2}O{sub 3} content and not containing Fe{sub 2}O{sub 3} were predominantly amorphous but subjected to devitrification under annealing. Addition of B{sub 2}O{sub 3} and partial Fe{sub 2}O{sub 3} substitution for Al{sub 2}O{sub 3} in the materials increases their resistance to devitrification whereas further substitution and NiO incorporation significantly increase the tendency to devitrification. FTIR spectra demonstrate changes in the structure of glassy materials caused by both structural variations in the anionic motif and occurrence of crystalline phases in the materials. According to Mössbauer spectroscopy data, iron in the glassy samples is present as octahedrally coordinated Fe{sup 3+} ions while in the partly devitrified samples iron is partitioned among vitreous and crystalline phases entering the vitreous phase mainly as Fe{sup 3+}O{sub 6} units and crystalline phases as major Fe{sup 3+} and minor Fe{sup 2+} ions in a magnetically ordered state and participating in a “fast” electronic exchange.

  15. Effects of alpha radiation on hardness and toughness of the borosilicate glass applied to radioactive wastes immobilization

    International Nuclear Information System (INIS)

    Prado, Miguel Oscar; Bernasconi, Norma B. Messi de; Bevilacqua, Arturo Miguel; Arribere, Maria Angelica; Heredia, Arturo D.; Sanfilippo, Miguel

    1999-01-01

    Borosilicate german glass SG7 samples, obtained by frit sintering, were irradiated with different fluences of thermal neutrons in the nucleus of a nuclear reactor. The nuclear reaction 10 B(n,α) 7 Li, where the 10 B isotope is one of the natural glass components, was used to generate alpha particles throughout the glass volume. The maximum alpha disintegration per unit volume achieved was equivalent to that accumulated in a borosilicate glass with nuclear wastes after 3.8 million years. Through Vickers indentations values for microhardness, stress for 50% fracture probability (Weibull statistics) and estimation of the toughness were obtained as a function of alpha radiation dose. Two counterbalanced effects were found: that due to the disorder created by the alpha particles in the glass and that due to the annealing during irradiation (temperature below 240 deg C). Considering the alpha radiation effect, glasses tend decrease Vickers hardness, and to increase thr 50% fracture probability stress with the dose increase. (author)

  16. Experimental Challenges and Successes in Measuring Aerosol Concentrations at Prototypic Spray Conditions Encountered at the Hanford Waste Treatment and Immobilization Plant - 13327

    Energy Technology Data Exchange (ETDEWEB)

    Bontha, J.R.; Gauglitz, P.A.; Kurath, D.E.; Adkins, H.E.; Enderlin, C.W.; Blanchard, J.; Daniel, R.C.; Song, C.; Schonewill, P.P.; Mahoney, L.A.; Buchmiller, W.C.; Boeringa, G.; Jenks, J. [Pacific Northwest National Laboratory, PO Box 999, Richland, Washington 99352 (United States)

    2013-07-01

    To date, majority of the work done on measuring aerosol releases from failure of process piping was done using simple Newtonian fluids and small engineered-nozzles that do not accurately represent the fluids and breaches postulated during accident analysis at the Hanford Waste Treatment and Immobilization Plant (WTP). In addition, the majority of the work conducted in this area relies on in-spray measurements that neglect the effect of splatter and do not yield any information regarding aerosol generation rates from this additional mechanism. In order to estimate aerosol generation rates as well as reduce the uncertainties in estimating the aerosol release fractions over a broad range of breaches, fluid properties and operating conditions encountered at the WTP, the Pacific Northwest National Laboratory (PNNL) has designed, commissioned, and tested two experimental test stands. The first test stand, referred to as the large-scale test stand, was designed specifically to measure aerosol concentrations and release fractions under prototypic conditions of flow and pressure for a range of breaches postulated in the hazard analysis for 0.076 m (3-inch) process pipes. However, the size of the large-scale test stand, anticipated fluid loss during a breach, experimental risks, and costs associated with hazardous chemical simulant testing limited the large-scale test stand utility to water and a few non-hazardous physical simulants that did not fully span the particle size and rheological properties of the fluids encountered at the WTP. Overcoming these limitations and extending the range of simulants used, required designing and building a smaller test stand, which was installed and operated in a fume hood. This paper presents some of the features of both test stands, the experimental challenges encountered, and successes in measuring aerosol concentration in both test stands over a range of test conditions. (authors)

  17. HLW immobilization in glass

    International Nuclear Information System (INIS)

    Leroy, P.; Jacquet-Francillon, N.; Runge, S.

    1992-01-01

    The immobilization of High Level Waste in glass in France is a long history which started as early as in the 1950's. More than 30 years of Research and Development have been invested in that field. Two industrial facilities are operating (AVM and R7) and a third one (T7), under cold testing, is planned to start active operation in the mid-92. While vitrification has been demonstrated to be an industrially mastered process, the question of the quality of the final waste product, i.e. the HLW glass, must be addressed. The scope of the present paper is to focus on the latter point from both standpoints of the R and D and of the industrial reality

  18. Production of Biodiesel Using Immobilized Lipase and the Characterization of Different Co-Immobilizing Agents and Immobilization Methods

    Directory of Open Access Journals (Sweden)

    Kang Zhao

    2016-08-01

    Full Text Available Lipase from Candida sp. 99–125 is widely employed to catalyzed transesterification and can be used for biodiesel production. In this study, the lipase was immobilized by combined adsorption and entrapment to catalyze biodiesel production from waste cooking oil (WCO via transesterification, and investigating co-immobilizing agents as additives according to the enzyme activity. The addition of the mixed co-immobilizing agents has positive effects on the activities of the immobilized lipase. Three different immobilizing methods were compared by the conversion ratio of biodiesel and structured by Atom Force Microscopy (AFM and Scanning Electron Microscopy (SEM, respectively. It was found that entrapment followed by adsorption was the best method. The effect of the co-immobilizing agent amount, lipase dosage, water content, and reuse ability of the immobilized lipase was investigated. By comparison with previous research, this immobilized lipase showed good reuse ability: the conversion ratio excesses 70% after 10 subsequent reactions, in particular, was better than Novozym435 and TLIM on waste cooking oil for one unit of lipase.

  19. Estudo da durabilidade de argamassas alternativas contendo resíduos Durability study of alternative mortars containing wastes

    Directory of Open Access Journals (Sweden)

    J. Farias Filho

    2011-12-01

    Full Text Available O interesse pela preservação dos recursos naturais e a busca pelo desenvolvimento sustentável tem despertado interesse de um grande número de pesquisadores visando obter materiais de construção não convencionais. Dentre essas possibilidades, destacam-se os resíduos produzidos pela construção civil e pela indústria de beneficiamento de granito, em virtude do volume gerado e os impactos ambientais do seu descarte inadequado. Este trabalho tem como objetivo estudar a durabilidade das argamassas alternativas incorporadas com resíduos de construção civil e da serragem do granito. Para quantificar a durabilidade dos materiais estudados, foram realizados ensaios de envelhecimento natural e ensaios de envelhecimento acelerado através de ciclos de molhagem e secagem, visando simular condições de utilização e avaliar sua degradação. Nos estudos da durabilidade foram utilizados traços de argamassas para alvenaria de elevação, com percentagens de substituição, em peso, da cal por resíduo de 30% e 50%. Os resultados para o envelhecimento natural evidenciaram o comprometimento da durabilidade do material após 60 dias com redução no comportamento mecânico, sendo os melhores resultados para os corpos-de-prova com 30% e 50% de resíduo de construção. No envelhecimento acelerado, para os tratamentos adotados, os resultados evidenciaram a ação da formação de silicoaluminatos de cálcio e potássio que favoreceram um aumento do comportamento mecânico e, além disso, não foi observado o comprometimento da sua durabilidade.The preservation of natural resources and search for sustainable development are increase in interest by the researchers with the aim to obtain knowledge of non conventional materials. Among these possibilities stand out the recycling of residues from civil construction and from granite industry, which are producers of large amount of residues and causing great environmental impacts. The aim of this work is to

  20. Immobilization of spent resin with epoxy resin

    International Nuclear Information System (INIS)

    Gultom, O.; Suryanto; Sayogo; Ramdan

    1997-01-01

    immobilization of spent resin using epoxy resin has been conducted. The spent resin was mixtured with epoxy resin in variation of concentration, i.e., 30, 40, 50, 60, 70 weight percent of spent resin. The mixture were pour into the plastic tube, with a diameter of 40 mm and height of 40 mm. The density, compressive strength and leaching rate were respectively measured by quanta chrome, paul weber apparatus and gamma spectrometer. The results showed that the increasing of waste concentration would be decreased the compressive strength, and increased density by immobilized waste. The leaching rate of 137 Cs from waste product was not detected in experiment (author)

  1. Status of plutonium ceramic immobilization processes and immobilization forms

    Energy Technology Data Exchange (ETDEWEB)

    Ebbinghaus, B.B.; Van Konynenburg, R.A. [Lawrence Livermore National Lab., CA (United States); Vance, E.R.; Jostsons, A. [Australian Nuclear Science and Technology Organization, Menai (Australia)] [and others

    1996-05-01

    Immobilization in a ceramic followed by permanent emplacement in a repository or borehole is one of the alternatives currently being considered by the Fissile Materials Disposition Program for the ultimate disposal of excess weapons-grade plutonium. To make Pu recovery more difficult, radioactive cesium may also be incorporated into the immobilization form. Valuable data are already available for ceramics form R&D efforts to immobilize high-level and mixed wastes. Ceramics have a high capacity for actinides, cesium, and some neutron absorbers. A unique characteristic of ceramics is the existence of mineral analogues found in nature that have demonstrated actinide immobilization over geologic time periods. The ceramic form currently being considered for plutonium disposition is a synthetic rock (SYNROC) material composed primarily of zirconolite (CaZrTi{sub 2}O{sub 7}), the desired actinide host phase, with lesser amounts of hollandite (BaAl{sub 2}Ti{sub 6}O{sub 16}) and rutile (TiO{sub 2}). Alternative actinide host phases are also being considered. These include pyrochlore (Gd{sub 2}Ti{sub 2}O{sub 7}), zircon (ZrSiO{sub 4}), and monazite (CePO{sub 4}), to name a few of the most promising. R&D activities to address important technical issues are discussed. Primarily these include moderate scale hot press fabrications with plutonium, direct loading of PuO{sub 2} powder, cold press and sinter fabrication methods, and immobilization form formulation issues.

  2. Status of plutonium ceramic immobilization processes and immobilization forms

    International Nuclear Information System (INIS)

    Ebbinghaus, B.B.; Van Konynenburg, R.A.; Vance, E.R.; Jostsons, A.

    1996-01-01

    Immobilization in a ceramic followed by permanent emplacement in a repository or borehole is one of the alternatives currently being considered by the Fissile Materials Disposition Program for the ultimate disposal of excess weapons-grade plutonium. To make Pu recovery more difficult, radioactive cesium may also be incorporated into the immobilization form. Valuable data are already available for ceramics form R ampersand D efforts to immobilize high-level and mixed wastes. Ceramics have a high capacity for actinides, cesium, and some neutron absorbers. A unique characteristic of ceramics is the existence of mineral analogues found in nature that have demonstrated actinide immobilization over geologic time periods. The ceramic form currently being considered for plutonium disposition is a synthetic rock (SYNROC) material composed primarily of zirconolite (CaZrTi 2 O 7 ), the desired actinide host phase, with lesser amounts of hollandite (BaAl 2 Ti 6 O 16 ) and rutile (TiO 2 ). Alternative actinide host phases are also being considered. These include pyrochlore (Gd 2 Ti 2 O 7 ), zircon (ZrSiO 4 ), and monazite (CePO 4 ), to name a few of the most promising. R ampersand D activities to address important technical issues are discussed. Primarily these include moderate scale hot press fabrications with plutonium, direct loading of PuO 2 powder, cold press and sinter fabrication methods, and immobilization form formulation issues

  3. Estudo morfométrico do músculo sóleo de ratos da linhagem wistar pós–imobilização articular = Morphometric study of post-joint immobilization of soleus muscle on wistar lineage rats

    Directory of Open Access Journals (Sweden)

    Sônia Maria Marques Gomes Bertolini

    2010-01-01

    Full Text Available Todos os tipos de imobilização contribuem para a atrofia muscular e, em apenas alguns dias, os músculos passam por diminuição de volume ou perda de função, conhecidos como atrofia. Assim, com uma ou duas semanas de imobilização, as atividades metabólicas são consideravelmente reduzidas e suas fibras musculares substituídas por tecido cicatricial fibroso denso. Dessa forma, este estudo teve como objetivo analisar o efeito da imobilização articular do músculo sóleo do membro posterior de ratos no perfil morfométrico, em diferentes períodos. Foram utilizados dez Rattus navergicus albinus machos, variedade Wistar, que foram divididos em dois grupos com cinco animais cada, sendo o primeiro grupo submetido à imobilização por sete dias e o segundo por 14 dias. O controle do experimento foi obtido a partir do membro contralateral direito do respectivo animal. A imobilização do membro posterior esquerdo foi por meio de uma órtese adaptada. A análise morfométrica do sóleo foi realizada por meio de cortes transversais não seriados de 5 μm de espessura. Foram analisadas, por meio das imagens obtidas, a área das fibras musculares, juntamente com a densidade do tecido conjuntivo, comparando-as ao Grupo-controle, referentes à perna contralateral. Com sete e 14 dias de imobilização, pode-se observar redução significativa (p All types of immobilization contribute to muscular atrophy and, in a few days, the muscles undergo volume reduction or loss of function, known as atrophy. Thus, with one or two weeks of immobilization, metabolic activities are considerably reduced and muscle fibers are replaced by dense fibrous scar tissue. This study has as objective to analyze the effect of joint immobilization of the soleus muscle of posterior members of rats on morphometric profile view, at periods of 7 and 14 days. Ten male Rattus navergicus albinus, Wistar variety, were used, separated into two groups of 5 animals each, with the first

  4. Immobilization needs and technology programs

    International Nuclear Information System (INIS)

    Gray, L.W.; Kan, T.; Shaw, H.; Armantrout, G.

    1995-01-01

    In the aftermath of the Cold War, the US and Russia agreed to large reductions in nuclear weapons. To aid in the selection of long-term management options, DOE has undertaken a multifaceted study to select options for storage and disposition of plutonium in keeping with US policy that plutonium must be subjected to the highest standards of safety, security, and accountability. One alternative being considered is immobilization. To arrive at a suitable immobilization form, we first reviewed published information on high-level waste immobilization technologies and identified 72 possible plutonium immobilization forms to be prescreened. Surviving forms were further screened using multi-attribute utility analysis to determine the most promising technology families. Promising immobilization families were further evaluated to identify chemical, engineering, environmental, safety, and health problems that remain to be solved prior to making technical decisions as to the viability of using the form for long- term disposition of plutonium. From this evaluation, a detailed research and development plan has been developed to provide answers to these remaining questions

  5. Immobilization of iodine in concrete

    Science.gov (United States)

    Clark, Walter E.; Thompson, Clarence T.

    1977-04-12

    A method for immobilizing fission product radioactive iodine recovered from irradiated nuclear fuel comprises combining material comprising water, Portland cement and about 3-20 wt. % iodine as Ba(IO.sub.3).sub.2 to provide a fluid mixture and allowing the fluid mixture to harden, said Ba(IO.sub.3).sub.2 comprising said radioactive iodine. An article for solid waste disposal comprises concrete prepared by this method. BACKGROUND OF THE INVENTION This invention was made in the course of, or under a contract with the Energy Research and Development Administration. It relates in general to reactor waste solidification and more specifically to the immobilization of fission product radioactive iodine recovered from irradiated nuclear fuel for underground storage.

  6. Immobilization of iodine in concrete

    International Nuclear Information System (INIS)

    Clark, W.E.; Thompson, C.T.

    1977-01-01

    A method for immobilizing fission product radioactive iodine recovered from irradiated nuclear fuel comprises combining material comprising water, Portland cement and about 3 to 20 wt percent iodine as Ba(IO 3 ) 2 to provide a fluid mixture and allowing the fluid mixture to harden, said Ba(IO 3 ) 2 comprising said radioactive iodine. An article for solid waste disposal comprises concrete prepared by this method. 10 claims, 2 figures

  7. Novel pectin-silica hybrids used for immobilization of Trichosporon cutaneum cells efficient in removal of Cadmium and Copper ions from waste water

    International Nuclear Information System (INIS)

    Georgieva, N.; Rangelova, N.; Peshev, D.; Nenkova, S.

    2011-01-01

    New silica hybrid materials containing tetramethyl siloxane (TMOS) as an inorganic precursor and apple pectin (AP) as an organic compound were prepared. The quantity of organic substance was 5 and 50 wt% AP. The amorphous state of the samples was proved by X-ray diffraction analyses (XRD). The Infrared scattering spectra (IR) showed characteristic peaks for SiO2 network, as well as for pectin. The synthesized hybrid materials were applied as matrices for cells immobilization by attachment and entrapment of the filamentous yeast Trichosporon cutaneum R57. This strain showed considerable ability to remove cadmium and copper ions from aqueous solutions. Regarding heavy metal biosorption capacity, the attachment was found to be superior compared to the entrapment method as a technique for biomass immobilization. (authors) Key words: biomaterials, composite materials, microstructure, sol-gel preparation

  8. estudo comparativo

    OpenAIRE

    Álvaro, Rocha

    2005-01-01

    O sítio web é um importante meio de comunicação e competitividade das universidades. Há diferentes razões para acreditar que os sítios possam variar entre universidades, mas contudo há certeza da existência de requisitos de qualidade comuns e transversais a todas. Este estudo visou avaliar, de forma quantitativa e objectiva, o cumprimento de requisitos de qualidade pelos sítios web das Universidades Portuguesas, com base nas características de alto nível da norma ISO 9126-1 que interessam aos...

  9. Study and development of phosphorus-32 source immobilized in polymer matrix for paraspinal and intracranial cancer treatment; Estudo e desenvolvimento de fonte de fosforo-32 imobilizado em matriz polimerica para tratamento de cancer paravertebral e intracranial

    Energy Technology Data Exchange (ETDEWEB)

    Benega, Marcos Antonio Gimenes

    2015-07-01

    The latest estimates of the World Health Organization show the occurrence of 14.1 million new cases of cancer in 2012. From these cases, 8.2 million will come to death. The paraspinal and intracranial tumors, also called central nervous system cancers, are originated in the brain, cranial nerves and meninges. A new brachytherapy modality began to be used in the last decade. In this procedure, flexible, polymeric plaques carrying phosphorus-32 are placed in contact or close to the tumor for treatment. This treatment has advantages over others because it applies a high dose rate in the tumor sparing healthy tissues. The production of these plaques is not well known, although there are satisfactory results in its use for the treatment of central nervous system cancers. This work carried out initial studies for the production of this type of polymer plaques for brachytherapy. The mechanical properties and immobilization capacity of radioactive material, from two commercial resins, epoxy and polyurethane, with or without the presence of polycarbonate as substrate were evaluated. Initial tests showed the use epoxy resin as the best alternative and the first prototypes and tests with use of it were made. The use of polycarbonate as a substrate was not required on one of the methodologies, facilitating the procedure but offering a lower security barrier. The tensile tests showed that addition of acid to the epoxy resin solution changed its mechanical properties, but there was a small improvement in flexibility. Adhesion tests showed better adhesion of the resin to the textured surface of the polycarbonate. The thermogravimetric analysis showed that the acid solution added to the resin structure is sealed even with temperature rises above 100°C. The epoxy resin used has the ability to incorporate the radioactive material in the form of acid solution and remain tight after wiping and immersion in hot liquid tests According to the results, the production of these plaques

  10. IN SITU LEAD IMMOBILIZATION BY APATITE

    Science.gov (United States)

    Lead contamination is of environmental concern due to its effect on human health. The purpose of this study was to develop a technology to immobilize Pb in situ in contaminated soils and wastes using apatite. Hydroxyapatite [Ca10(PO4)6(O...

  11. Bioreduction and immobilization of uranium in situ: a case study at a USA Department of Energy radioactive waste site, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Wu, Weimin; Carley, Jack M.; Watson, David B.; Gu, Baohua; Brooks, Scott C.; Kelly, Shelly D.; Kemner, Kenneth M.; Van Nostrand, Joy; Wu, Liyou; Zhou, Jizhong; Luo, Jian; Cardenas, Erick; Fields, Matthew Wayne; Marsh, Terence; Tiedje, James; Green, Stefan; Kostka, Joel; Kitanidis, Peter K.; Jardine, Philip; Criddle, Craig

    2011-01-01

    Bioremediation of uranium contaminated groundwater was tested by delivery of ethanol as an electron donor source to stimulate indigenous microbial bioactivity for reduction and immobilization of uranium in situ, followed by tests of stability of uranium sequestration in the bioreduced area via delivery of dissolved oxygen or nitrate at the US Department of energy's Integrated Field Research Challenge site located at Oak Ridge, Tennessee, USA. After long term treatment that spanned years, uranium in groundwater was reduced from 40-60 mg · L -1 to -1 , below the USA EPA standard for drinking water. The bioreduced uranium was stable under anaerobic or anoxic conditions, but addition of DO and nitrate to the bioreduced zone caused U remobilization. The change in the microbial community and functional microorganisms related to uranium reduction and oxidation were characterized. The delivery of ethanol as electron donor stimulated the activities of indigenous microorganisms for reduction of U(VI) to U(IV). Results indicated that the immobilized U could be partially remobilized by D0 and nitrate via microbial activity. An anoxic environmental condition without nitrate is essential to maintain the stability of bioreduced uranium.

  12. Effect of the nature of alkali and alkaline-earth oxides on the structure and crystallization of an alumino-borosilicate glass developed to immobilize highly concentrated nuclear waste solutions

    International Nuclear Information System (INIS)

    Quintas, A.; Caurant, D.; Majerus, O.; Charpentier, T.; Dussossoy, J.L.

    2008-01-01

    A complex rare-earth rich alumino-borosilicate glass has been proved to be a good candidate for the immobilization of new high level radioactive wastes. A simplified seven-oxides composition of this glass was selected for this study. In this system, sodium and calcium cations were supposed in other works to simulate respectively all the other alkali (R + = Li + , Rb + , Cs + ) and alkaline-earth (R 2+ = Sr 2+ , Ba 2+ ) cations present in the complex glass composition. Moreover, neodymium or lanthanum are used here to simulate all the rare-earths and actinides occurring in waste solutions. In order to study the impact of the nature of R + and R 2+ cations on both glass structure and melt crystallization tendency during cooling, two glass series were prepared by replacing either Na + or Ca 2+ cations in the simplified glass by respectively (Li + , K + , Rb + , Cs + ) or (Mg 2+ , Sr 2+ , Ba 2+ ) cations. From these substitutions, it was established that alkali ions are preferentially involved in the charge compensation of (AlO 4 ) - entities in the glass network comparatively to alkaline-earth ions. The glass compositions containing calcium give way to the crystallization of an apatite silicate phase bearing calcium and rare-earth ions. The melt crystallization tendency during cooling strongly varies with the nature of the alkaline-earth. (authors)

  13. Immobilization and Limited Reoxidation of Technetium-99 by Fe(II)-Goethite

    Energy Technology Data Exchange (ETDEWEB)

    Um, Wooyong; Chang, Hyun-shik; Icenhower, Jonathan P.; Qafoku, Nikolla; Smith, Steven C.; Serne, R. Jeffrey; Buck, Edgar C.; Kukkadapu, Ravi K.; Bowden, Mark E.; Westsik, Joseph H.; Lukens, Wayne W.

    2010-09-30

    This report summarizes the methodology used to test the sequestration of technetium-99 present in both deionized water and simulated Hanford Tank Waste Treatment and Immobilization Plant waste solutions.

  14. The case study of management of solid wastes in a petroleum industry; O estudo de caso do gerenciamento de residuos solidos em uma refinaria de petroleo

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Lizabela Souza de [Universidade Federal, Rio de Janeiro, RJ (Brazil). Escola de Quimica]. E-mail: lizabela@eq.ufrj.br; Nicolaiewsky, Elioni [Universidade Federal, Rio de Janeiro, RJ (Brazil). Escola de Quimica. Dept. de Engenharia Quimica]. E-mail: elioni@eq.ufrj.br; Freire, Denize D.C. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Escola de Quimica. Dept. de Engenharia Bioquimica]. E-mail: denize@eq.ufrj.br

    2003-07-01

    Crude oil refining is an industrial activity known as very pollutant, as all other activities of the petroleum industry, regarding either the volume or the concentration of the resides involved, thus generating emissions, effluents and solid wastes. The aim of the present work is to study solid waste management of a certain petroleum refinery, located in Rio de Janeiro. On the solid wastes management of that refinery, the following aspects were considered: origin and period of generation, conditioning, storage, transportation, treatment and final disposal. After listing all the resides and through analysis of the industrial wastes (norms, terms, inventory), the industrial process and office routines were then analyzed. The solid wastes were divided in two categories: industrial and administrative wastes. As far as destination is concerned, resides classified as Class I are either co-processed or incinerated, while Class II and Class III wastes, when not recycled, are sent to industrial or sanitary landfill. Finally, after analyzing the wastes management of the refinery, it has been proposed a plan of achievements in order to enhance the environmental goal of the refinery. (author)

  15. Sodium aluminum-iron phosphate glass-ceramics for immobilization of lanthanide oxide wastes from pyrochemical reprocessing of spent nuclear fuel

    Science.gov (United States)

    Stefanovsky, S. V.; Stefanovsky, O. I.; Kadyko, M. I.; Nikonov, B. S.

    2018-03-01

    Sodium aluminum (iron) phosphate glass ceramics containing of up to 20 wt.% rare earth (RE) oxides simulating pyroprocessing waste were produced by melting at 1250 °C followed by either quenching or slow cooling to room temperature. The iron-free glass-ceramics were composed of major glass and minor phosphotridymite and monazite. The iron-bearing glass-ceramics were composed of major glass and minor monazite and Na-Al-Fe orthophosphate at low waste loadings (5-10 wt.%) and major orthophosphate and minor monazite as well as interstitial glass at high waste loadings (15-20 wt.%). Slowly cooled samples contained higher amount of crystalline phases than quenched ones. Monazite is major phase for REs. Leach rates from the materials of major elements (Na, Al, Fe, P) are 10-5-10-7 g cm-2 d-1, RE elements - lower than 10-5 g cm-2 d-1.

  16. Remote handling in the Plutonium Immobilization Project: Puck handling

    International Nuclear Information System (INIS)

    Brault, J.R.

    2000-01-01

    Since the break up of the Soviet Union at the end of the Cold War, the US and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

  17. Immobilized enzymes and cells

    Energy Technology Data Exchange (ETDEWEB)

    Bucke, C; Wiseman, A

    1981-04-04

    This article reviews the current state of the art of enzyme and cell immobilization and suggests advances which might be made during the 1980's. Current uses of immobilized enzymes include the use of glucoamylase in the production of glucose syrups from starch and glucose isomerase in the production of high fructose corn syrup. Possibilities for future uses of immobilized enzymes and cells include the utilization of whey and the production of ethanol.

  18. Plutonium Immobilization Can Loading Conceptual Design

    International Nuclear Information System (INIS)

    Kriikku, E.

    1999-01-01

    'The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with glass for permanent storage. This report discusses the Plutonium Immobilization can loading conceptual design and includes a process block diagram, process description, preliminary equipment specifications, and several can loading issues. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas.'

  19. Plutonium Immobilization Can Loading Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Kriikku, E.

    1999-05-13

    'The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with glass for permanent storage. This report discusses the Plutonium Immobilization can loading conceptual design and includes a process block diagram, process description, preliminary equipment specifications, and several can loading issues. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas.'

  20. Ceramics in nuclear waste management

    Energy Technology Data Exchange (ETDEWEB)

    Chikalla, T D; Mendel, J E [eds.

    1979-05-01

    Seventy-three papers are included, arranged under the following section headings: national programs for the disposal of radioactive wastes, waste from stability and characterization, glass processing, ceramic processing, ceramic and glass processing, leaching of waste materials, properties of nuclear waste forms, and immobilization of special radioactive wastes. Separate abstracts were prepared for all the papers. (DLC)

  1. Hot isostatically-pressed aluminosilicate glass-ceramic with natural crystalline analogues for immobilizing the calcined high-level nuclear waste at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Raman, S.

    1993-12-01

    The additives Si, Al, MgO, P 2 O 5 were mechanically blended with fluorinelsodium calcine in varying proportions. The batches were vacuum sealed in stainless steel canisters and hot isostatically pressed at 20,000 PSI and 1000 C for 4 hours. The resulting suite of glass-ceramic waste forms parallels the natural rocks in microstructural and compositional heterogeneity. Several crystalline phases ar analogous in composition and structure to naturally occurring minerals. Additional crystalline phases are zirconia and Ca-Mg borate. The glasses are enriched in silica and alumina. Approximately 7% calcine elements occur dissolved in this glass and the total glass content in the waste forms averages 20 wt%. The remainder of the calcine elements are partitioned into crystalline phases at 75 wt% calcine waste loading. The waste forms were tested for chemical durability in accordance with the MCC1-test procedure. The leach rates are a function of the relative proportions of additives and calcine, which in turn influence the composition and abundances of the glass and crystalline phases. The DOE leach rate criterion of less than 1 g/m 2 -day is met by all the elements B, Cs and Na are increased by lowering the melt viscosity. This is related to increased crystallization or devitrification with increases in MgO addition. This exploratory work has shown that the increases in waste loading occur by preferred partitioning of the calcine components among crystalline and glass phases. The determination of optimum processing parameters in the form of additive concentration levels, homogeneous blending among the components, and pressure-temperature stabilities of phases must be continued to eliminate undesirable effects of chemical composition, microstructure and glass devitrification

  2. Design-Only Conceptual Design Report: Plutonium Immobilization Plant

    International Nuclear Information System (INIS)

    DiSabatino, A.; Loftus, D.

    1999-01-01

    This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant

  3. Transuranic waste management program waste form development

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  4. Nuclear waste management. Quarterly progress report, January-March 1980

    Energy Technology Data Exchange (ETDEWEB)

    Platt, A.M.; Powell, J.A. (comps.)

    1980-06-01

    Reported are: high-level waste immobilization, alternative waste forms, nuclear waste materials characterization, TRU waste immobilization, TRU waste decontamination, krypton solidification, thermal outgassing, iodine-129 fixation, unsaturated zone transport, well-logging instrumentation development, mobile organic complexes of fission products, waste management system and safety studies, assessment of effectiveness of geologic isolation systems, waste/rock interactions, engineered barriers, criteria for defining waste isolation, and spent fuel and pool component integrity. (DLC)

  5. Immobilization of high activity nuclear wastes in sintered glass. Fabrication of blocks at semi-industrial scale by hot pressing technique

    International Nuclear Information System (INIS)

    Russo, D.O.; Messi, N.B.; Riquelme, R.; Sterba, M.E.; Audero, M.A.

    1990-01-01

    The sintering process under glass pressure has been studied as an alternative of melting with the aim of obtaining a monolytic material apt to preserve the high activity nuclear wastes. Different properties of the products obtained have been evaluated where the material is selected on the basis of the results attained. The purpose of this work is the equipment development and the process adjusting for the blocks obtainment. (Author) [es

  6. Immobilization of technetium and nitrate in cement-based materials

    International Nuclear Information System (INIS)

    Tallent, O.K.; McDaniel, E.W.; Del Cul, G.D.; Dodson, K.E.; Trotter, D.R.

    1987-01-01

    The leachabilities of technetium and nitrate wastes immobilized in cement-based grouts have been investigated. Factors found to affect the leachabilities include grout mix ratio, grout fluid density, dry solid blend composition, and waste concentration. 10 refs., 7 figs., 3 tabs

  7. Production and characterization of red mud based on glasses for the immobilization of nuclear wastes; Obtencao e caracterizacao de vidros a base de lama vermelha visando a imobilizacao de rejeitos nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Heveline

    2015-07-01

    Glasses based on red mud, a residual material from bauxite processing, were developed and characterized in this work. In order to promote its use, a minimum 60 wt% of red mud was used in the production of the glasses. According to XRD results, materials containing considerable amorphous phases were produced when using red mud as raw material. These amorphous phases were observed even though crystalline phases associated to Fe coming from the red mud itself were present. The material denominated 60L40S, which has a nominal composition of 60 wt% red mud showed the best properties comparing with the others compositions studied. However, these materials presented a high melting temperature. Changes in the composition of this material were made with the objective of lowering this temperature. Results indicated that the changes made to the material were successful in the reduction of the melting temperature. However, a reduction in the chemical properties of the resulting material was observed. Elements usually found in the chemical composition of nuclear wastes were added to the glasses produced. It was done with the objective of determining the effect of these elements on the chemical and physical properties of the red mud based glasses obtained. It was found that it was possible to add up to 15 wt% of these elements to the materials produced. The addition of these simulant materials promoted a reduction in the melting temperature of the resulting material. Above 15 wt%, the added elements precipitate in the structure of the resulting material. Even though the reduction in the chemical durability of the 60L40S material when simulant elements were added, it was observed that this material contained the simulant elements confined in its structure when in contact with water. This is a promising result, since it indicates that the 60L40S has the potential to immobilize elements from nuclear wastes . (author)

  8. Particular provisions applicable to the production, inspection, treatment, packaging and interim storage of wastes immobilized in cement, resulting from the reprocessing of fuels irradiated in pressurized light water reactors

    International Nuclear Information System (INIS)

    1985-02-01

    The Fundamental Safety Rules applicable to certain types of nuclear installation are intended to clarify the conditions of which observance, for the type of installation concerned and for the subject that they deal with, is considered as equivalent to compliance with regulatory French technical practice. These Rules should facilitate safety analysises and the clear understanding between persons interested in matters related to nuclear safety. They in no way reduce the operator's liability and pose no obstacle to statutory provisions in force. For any installation to which a Fundamental Safety Rule applies according to the foregoing paragraph, the operator may be relieved from application of the Rule if he shows proof that the safety objectives set by the Rule are attained by other means that he proposes within the framework of statutory procedures. Furthermore, the Central Service for the Safety of Nuclear Installations reserves the right at all times to alter any Fundamental Safety Rule, as required, should it deem this necessary, while specifying the applicability conditions. This rule is intended to stipulate the specific provisions applicable to the production, inspection, treatment, packaging and interim storage of the wastes, resulting from the reprocessing of fuels irradiated in a PWR and immobilized in cement

  9. Immobilization of carbon-14 from reactor graphite waste by use of self-sustaining reaction in the C-Al-TiO2 system

    International Nuclear Information System (INIS)

    Karlina, O.K.; Klimov, V.L.; Ojovan, M.I.; Pavlova, G.Yu.; Dmitriev, S.A.; Yurchenko, A.Yu.

    2005-01-01

    As a result of long-term neutron irradiation, the long-lived 14 C is produced in the reactor graphite. The exothermic self-sustaining reaction 3C(graphite) + 4Al + 3TiO 2 = 3TiC + 2Al 2 O 3 was proposed for processing of such waste. In doing so, the carbon, including the 14 C, is chemically bound in the stable TiC. The reaction products in the C-Al-TiO 2 system were investigated both by thermodynamic simulation and experimentally in the course of this work

  10. Crystalline phase, microstructure, and aqueous stability of zirconolite-barium borosilicate glass-ceramics for immobilization of simulated sulfate bearing high-level liquid waste

    Science.gov (United States)

    Wu, Lang; Xiao, Jizong; Wang, Xin; Teng, Yuancheng; Li, Yuxiang; Liao, Qilong

    2018-01-01

    The crystalline phase, microstructure, and aqueous stability of zirconolite-barium borosilicate glass-ceramics with different content (0-30 wt %) of simulated sulfate bearing high-level liquid waste (HLLW) were evaluated. The sulfate phase segregation in vitrification process was also investigated. The results show that the glass-ceramics with 0-20 wt% of HLLW possess mainly zirconolite phase along with a small amount baddeleyite phase. The amount of perovskite crystals increases while the amount of zirconolite crystals decreases when the HLLW content increases from 20 to 30 wt%. For the samples with 20-30 wt% HLLW, yellow phase was observed during the vitrification process and it disappeared after melting at 1150 °C for 2 h. The viscosity of the sample with 16 wt% HLLW (HLLW-16) is about 27 dPa·s at 1150 °C. The addition of a certain amount (≤20 wt %) of HLLW has no significant change on the aqueous stability of glass-ceramic waste forms. After 28 days, the 90 °C PCT-type normalized leaching rates of Na, B, Si, and La of the sample HLLW-16 are 7.23 × 10-3, 1.57 × 10-3, 8.06 × 10-4, and 1.23 × 10-4 g·m-2·d-1, respectively.

  11. Waste management: products and services

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    A number of products and services related to radioactive waste management are described. These include: a portable cement solidification system for waste immobilization; spent fuel storage racks; storage and transport flasks; an on-site low-level waste storage facility; supercompactors; a mobile waste retrieval and encapsulation plant; underwater crushers; fuel assembly disposal; gaseous waste management; environmental restoration and waste management services; a waste treatment consultancy. (UK)

  12. Production of cellulase from immobilized Trichoderma reesei

    International Nuclear Information System (INIS)

    Kasai, Noboru; Tamada, Masao; Kumakura, Minoru

    1989-05-01

    This report completed the results that obtained on the study of the enzyme activity in the culture of immobilized Trichoderma reesei cells in flask scale (100ml) and bench scale (30l). In the flask scale culture, the batch and repeated batch culture were carried out, and in the bench scale culture, the batch, repeated batch and continuous culture were done by using a culture equipment that is an unit process of the bench scale test plant for saccharification of cellulosic wastes. The enzyme activity of the immobilized cells was higher than that of the intact cells in the flask scale culture and it was confirmed that the enzyme activity was not decreased on the repeated batch culture of six times even. In the bench scale culture, it was found that a optimum culture condition of the immobilized cells was not different from that of the free cells and the immobilized cells gave the enzyme solution with a high enzyme activity in the culture condition of 450rpm stirring speed and air supply of 0.1v/v/m above. The technique of the repeated batch and continuous culture for long times in bench scale without contamination was established. The enzyme activity of the immobilized cells in continuous culture became to be 85 % to that in batch culture and it was found that the enzyme solution with high enzyme activity was continuously obtained in the continuous culture for long times. (author)

  13. Plutonium immobilization in glass and ceramics

    International Nuclear Information System (INIS)

    Knecht, D.A.; Murphy, W.M.

    1996-01-01

    The Materials Research Society Nineteenth Annual Symposium on the Scientific Basis for Nuclear Waste Management was held in Boston on November 27 to December 1, 1995. Over 150 papers were presented at the Symposium dealing with all aspects of nuclear waste management and disposal. Fourteen oral sessions and on poster session included a Plenary session on surplus plutonium dispositioning and waste forms. The proceedings, to be published in April, 1996, will provide a highly respected, referred compilation of the state of scientific development in the field of nuclear waste management. This paper provides a brief overview of the selected Symposium papers that are applicable to plutonium immobilization and plutonium waste form performance. Waste forms that were described at the Symposium cover most of the candidate Pu immobilization options under consideration, including borosilicate glass with a melting temperature of 1150 degrees C, a higher temperature (1450 degrees C) lanthanide glass, single phase ceramics, multi-phase ceramics, and multi-phase crystal-glass composites (glass-ceramics or slags). These Symposium papers selected for this overview provide the current status of the technology in these areas and give references to the relevant literature

  14. Plutonium immobilization in glass and ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Knecht, D.A. [Lockheed Martin Idaho Technologies, Idaho Falls (United States); Murphy, W.M. [Southwest Research Institute, San Antonio, TX (United States)

    1996-05-01

    The Materials Research Society Nineteenth Annual Symposium on the Scientific Basis for Nuclear Waste Management was held in Boston on November 27 to December 1, 1995. Over 150 papers were presented at the Symposium dealing with all aspects of nuclear waste management and disposal. Fourteen oral sessions and on poster session included a Plenary session on surplus plutonium dispositioning and waste forms. The proceedings, to be published in April, 1996, will provide a highly respected, referred compilation of the state of scientific development in the field of nuclear waste management. This paper provides a brief overview of the selected Symposium papers that are applicable to plutonium immobilization and plutonium waste form performance. Waste forms that were described at the Symposium cover most of the candidate Pu immobilization options under consideration, including borosilicate glass with a melting temperature of 1150 {degrees}C, a higher temperature (1450 {degrees}C) lanthanide glass, single phase ceramics, multi-phase ceramics, and multi-phase crystal-glass composites (glass-ceramics or slags). These Symposium papers selected for this overview provide the current status of the technology in these areas and give references to the relevant literature.

  15. Wastes

    International Nuclear Information System (INIS)

    Bovard, Pierre

    The origin of the wastes (power stations, reprocessing, fission products) is determined and the control ensuring the innocuity with respect to man, public acceptance, availability, economics and cost are examined [fr

  16. Enzyme Immobilization: An Overview on Methods, Support Material, and Applications of Immobilized Enzymes.

    Science.gov (United States)

    Sirisha, V L; Jain, Ankita; Jain, Amita

    Immobilized enzymes can be used in a wide range of processes. In recent years, a variety of new approaches have emerged for the immobilization of enzymes that have greater efficiency and wider usage. During the course of the last two decades, this area has rapidly expanded into a multidisciplinary field. This current study is a comprehensive review of a variety of literature produced on the different enzymes that have been immobilized on various supporting materials. These immobilized enzymes have a wide range of applications. These include applications in the sugar, fish, and wine industries, where they are used for removing organic compounds from waste water. This study also reviews their use in sophisticated biosensors for metabolite control and in situ measurements of environmental pollutants. Immobilized enzymes also find significant application in drug metabolism, biodiesel and antibiotic production, bioremediation, and the food industry. The widespread usage of immobilized enzymes is largely due to the fact that they are cheaper, environment friendly, and much easier to use when compared to equivalent technologies. © 2016 Elsevier Inc. All rights reserved.

  17. US Department of Energy Plutonium Stabilization and Immobilization Workshop, December 12-14, 1995: Final proceedings

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-05-01

    The purpose of the workshop was to foster communication within the technical community on issues surrounding stabilization and immobilization of the Department`s surplus plutonium and plutonium- contaminated wastes. The workshop`s objectives were to: build a common understanding of the performance, economics and maturity of stabilization and immobilization technologies; provide a system perspective on stabilization and immobilization technology options; and address the technical issues associated with technologies for stabilization and immobilization of surplus plutonium and plutonium- contaminated waste. The papers presented during this workshop have been indexed separately.

  18. Remote handling in the Plutonium Immobilization Project: Plutonium conversion and first stage immobilization

    International Nuclear Information System (INIS)

    Brault, J.R.

    2000-01-01

    Since the break up of the Soviet Union at the end of the Cold War, the United States and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

  19. Nuclear waste management. Quarterly progress report, October-December 1979

    Energy Technology Data Exchange (ETDEWEB)

    Platt, A.M.; Powell, J.A. (comps.)

    1980-04-01

    Progress and activities are reported on the following: high-level waste immobilization, alternative waste forms, nuclear waste materials characterization, TRU waste immobilization programs, TRU waste decontamination, krypton solidification, thermal outgassing, iodine-129 fixation, monitoring of unsaturated zone transport, well-logging instrumentation development, mobile organic complexes of fission products, waste management system and safety studies, assessment of effectiveness of geologic isolation systems, waste/rock interactions technology, spent fuel and fuel pool integrity program, and engineered barriers. (DLC)

  20. Plutonium Immobilization Can Loading Concepts

    International Nuclear Information System (INIS)

    Kriikku, E.; Ward, C.; Stokes, M.; Randall, B.; Steed, J.; Jones, R.; Hamilton, L.; Rogers, L.; Fiscus, J.; Dyches, G.

    1998-05-01

    The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with glass for permanent storage. This report discusses five can loading conceptual designs and the lists the advantages and disadvantages for each concept. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas. The can loading welder and cutter are very similar to the existing Savannah River Site (SRS) FB-Line bagless transfer welder and cutter and thus they are a low priority development item

  1. Field demonstration of in situ treatment of buried low-level radioactive solid waste with caustic soda and soda ash to immobilize 90Sr

    International Nuclear Information System (INIS)

    Spalding, B.P.

    1984-02-01

    A low-level radioactive solid waste disposal trench was injected on four occasions with solutions of caustic soda, soda ash, caustic soda, and lime/soda ash, respectively. Because investigations had indicated that 90 Sr could be coprecipitated with soil calcium carbonate by treatment with soda ash, this demonstration was undertaken as a test of its technical feasibility. After concentrations of 90 Sr and water hardness decreased within the intratrench monitoring wells; one well at the foot of the trench decreased from over 100 to a persistent level of less than 10 kBq of 90 Sr per liter. Recharge of 90 Sr from the trench to a sump immediately below was reduced by about 90%. Water hardness and 90 Sr concentrations were strongly correlated through time within each monitoring well, indicating that 90 Sr behaved as a tracer for soil calcium and magnesium. The disappearance of 90 Sr from the trench water, therefore, was an in situ water softening. Soil samples retrieved from the trench indicated that as much as 98% of the total 90 Sr was present as a coprecipitate with calcium carbonate. The hydrologic characterization of this trench indicated an average void space of 41% and an average trench-wall hydraulic conductivity of 3.4 x 10 -7 m/s. Sampling of the trench's discharge contamination plume indicated that it had resulted from a combination of subsurface seepage and bathtub overflow during infrequent periods of intense precipitation. A generic assessment of soda ash treatment indicated that treatment would be most effective for soils of high cation exchange capacity with either low ( 80%) basic cation saturation of that cation exchange capacity

  2. The thorium phosphate diphosphate as matrix for radioactive waste conditioning: radionuclide immobilization and behavior under irradiation; Le phosphate diphosphate de thorium, matrice pour le conditionnement des dechets radioactifs: immobilisation de radionucleides, comportement sous irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Pichot, Erwan [Inst. de Physique Nucleaire, Paris-11 Univ., 91 - Orsay (France)

    1999-04-13

    The aim of this work was to perform successively the decontamination of liquid solutions and the final immobilization of radionuclide storage using the same matrix. For this, thorium phosphate-diphosphate (TPD) of the formula Th{sub 4}P{sub 6}O{sub 23}, is proposed as a very resistant to water corrosion matrix. A new compound, thorium phosphate hydrogeno-phosphate (TPHP) of the formula Th{sub 2}(PO{sub 4}){sub 2}(HPO{sub 4}), nH{sub 2}O with n=3-7 was synthesized and characterized. Heated at 1100 deg.C it is transformed into the TDP. Ion exchange properties of TPHP were investigated. The exchange yields of imponderable caesium, strontium and americium ion onto TPHP (NaNO{sub 3} 0.1 M media at pH=6) are equal to 60% for the first one and 100% for the two others. The results interpreted in terms of ion-exchange led to determine selectivity coefficient values for each cation and suggested that only hydrated ions are exchanged. While the TPD is proposed for the high level nuclear waste storage, the irradiation effects, particularly structural modifications were studied using both {gamma} irradiation and charged particle irradiation. ESR and TL methods were carried out in order to identify radicals created during gamma radiation exposure. Correlation between ESR and TL experiments performed at room temperature clearly show three of PO{sub 3}{sup 2-} species and one POO{center_dot} species of free radicals. We have shown that Au-ion irradiation in the range of MeV energy involved TPD structure and chemical modifications. Important sputtering was interpreted in terms of local thermal chemical decomposition. We have shown, at room temperature, that the amorphization dose for heavy ion irradiation is between 0.1 to 0.4 dpa. (author) 146 refs., 46 figs., 21 tabs.

  3. 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

  4. Options for the Separation and Immobilization of Technetium

    International Nuclear Information System (INIS)

    Serne, R Jeffrey; Crum, Jarrod V.; Riley, Brian J.; Levitskaia, Tatiana G.

    2016-01-01

    Among radioactive constituents present in the Hanford tank waste, technetium-99 (Tc) presents a unique challenge in that it is significantly radiotoxic, exists predominantly in the liquid low-activity waste (LAW), and has proven difficult to effectively stabilize in a waste form for ultimate disposal. Within the Hanford Tank Waste Treatment and Immobilization Plant, the LAW fraction will be converted to a glass waste form in the LAW vitrification facility, but a significant fraction of Tc volatilizes at the high glass-melting temperatures and is captured in the off-gas treatment system. This necessitates recycle of the off-gas condensate solution to the LAW glass melter feed. The recycle process is effective in increasing the loading of Tc in the immobilized LAW (ILAW), but it also disproportionately increases the sulfur and halides in the LAW melter feed, which have limited solubility in the LAW glass and thus significantly reduce the amount of LAW (glass waste loading) that can be vitrified and still maintain good waste form properties. This increases both the amount of LAW glass and either the duration of the LAW vitrification mission or requires the need for supplemental LAW treatment capacity. Several options are being considered to address this issue. Two approaches attempt to minimize the off-gas recycle by removing Tc at one of several possible points within the tank waste processing flowsheet. The separated Tc from these two approaches must then be dispositioned in a manner such that the Tc can be safely disposed. Alternative waste forms that do not have the Tc volatility issues associated with the vitrification process are being sought for immobilization of Tc for subsequent storage and disposal. The first objective of this report is to provide insights into the compositions and volumes of the Tc-bearing waste streams including the ion exchange eluate from processing LAW and the off-gas condensate from the melter. The first step to be assessed will be the

  5. Options for the Separation and Immobilization of Technetium

    Energy Technology Data Exchange (ETDEWEB)

    Serne, R Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Levitskaia, Tatiana G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-30

    Among radioactive constituents present in the Hanford tank waste, technetium-99 (Tc) presents a unique challenge in that it is significantly radiotoxic, exists predominantly in the liquid low-activity waste (LAW), and has proven difficult to effectively stabilize in a waste form for ultimate disposal. Within the Hanford Tank Waste Treatment and Immobilization Plant, the LAW fraction will be converted to a glass waste form in the LAW vitrification facility, but a significant fraction of Tc volatilizes at the high glass-melting temperatures and is captured in the off-gas treatment system. This necessitates recycle of the off-gas condensate solution to the LAW glass melter feed. The recycle process is effective in increasing the loading of Tc in the immobilized LAW (ILAW), but it also disproportionately increases the sulfur and halides in the LAW melter feed, which have limited solubility in the LAW glass and thus significantly reduce the amount of LAW (glass waste loading) that can be vitrified and still maintain good waste form properties. This increases both the amount of LAW glass and either the duration of the LAW vitrification mission or requires the need for supplemental LAW treatment capacity. Several options are being considered to address this issue. Two approaches attempt to minimize the off-gas recycle by removing Tc at one of several possible points within the tank waste processing flowsheet. The separated Tc from these two approaches must then be dispositioned in a manner such that the Tc can be safely disposed. Alternative waste forms that do not have the Tc volatility issues associated with the vitrification process are being sought for immobilization of Tc for subsequent storage and disposal. The first objective of this report is to provide insights into the compositions and volumes of the Tc-bearing waste streams including the ion exchange eluate from processing LAW and the off-gas condensate from the melter. The first step to be assessed will be the

  6. Integrated development and testing plan for the plutonium immobilization project

    International Nuclear Information System (INIS)

    Kan, T.

    1998-01-01

    This integrated plan for the DOE Office of Fissile Materials Disposition (MD) describes the technology development and major project activities necessary to support the deployment of the immobilization approach for disposition of surplus weapons-usable plutonium. The plan describes details of the development and testing (D and T) tasks needed to provide technical data for design and operation of a plutonium immobilization plant based on the ceramic can-in-canister technology (''Immobilization Fissile Material Disposition Program Final Immobilization Form Assessment and Recommendation'', UCRL-ID-128705, October 3, 1997). The plan also presents tasks for characterization and performance testing of the immobilization form to support a repository licensing application and to develop the basis for repository acceptance of the plutonium form. Essential elements of the plant project (design, construction, facility activation, etc.) are described, but not developed in detail, to indicate how the D and T results tie into the overall plant project. Given the importance of repository acceptance, specific activities to be conducted by the Office of Civilian Radioactive Waste Management (RW) to incorporate the plutonium form in the repository licensing application are provided in this document, together with a summary of how immobilization D and T activities provide input to the license activity. The ultimate goal of the Immobilization Project is to develop, construct, and operate facilities that will immobilize from about 18 to 50 tonnes (MT) of U.S. surplus weapons usable plutonium materials in a manner that meets the ''spent fuel'' standard (Fissile Materials Storage and Disposition Programmatic Environmental Impact Statement Record of Decision, ''Storage and Disposition Final PEIS'', issued January 14, 1997, 62 Federal Register 3014) and is acceptable for disposal in a geologic repository. In the can-in-canister technology, this is accomplished by encapsulating the

  7. Immobilization of microorganisms. Part 1. Preparation of immobilized Lactobacillus bulgaricus

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K H

    1981-01-01

    The immobilization of Lactobacillus bulgaricus on polyacrylamide and on alginate beads was investigated. The most active immobilized cells were obtained by entrapment in Ca alginate beads. These immobilized microbial cells, when introduced into 4.5% lactose solution and whey solution showed maximum relative activity of 28% for lactose and 18% for whey compared to free cells.

  8. Mathematical model for choosing the nuclear safe matrix compositions for fissile material immobilization

    International Nuclear Information System (INIS)

    Gorshtein, A.I.; Matyunin, Yu.I.; Poluehktov, P.P.

    2000-01-01

    A mathematical model is proposed for preliminary choice of the nuclear safe matrix compositions for fissile material immobilization. The IBM PC computer software for nuclear safe matrix composition calculations is developed. The limiting concentration of fissile materials in the some used and perspective nuclear safe matrix compositions for radioactive waste immobilization is calculated [ru

  9. Limb immobilization and corticobasal syndrome.

    Science.gov (United States)

    Graff-Radford, Jonathan; Boeve, Bradley F; Drubach, Daniel A; Knopman, David S; Ahlskog, J Eric; Golden, Erin C; Drubach, Dina I; Petersen, Ronald C; Josephs, Keith A

    2012-12-01

    Recently, we evaluated two patients with corticobasal syndrome (CBS) who reported symptom onset after limb immobilization. Our objective was to investigate the association between trauma, immobilization and CBS. The charts of forty-four consecutive CBS patients seen in the Mayo Clinic Alzheimer Disease Research Center were reviewed with attention to trauma and limb immobilization. 10 CBS patients (23%) had immobilization or trauma on the most affected limb preceding the onset or acceleration of symptoms. The median age at onset was 61. Six patients manifested their first symptoms after immobilization from surgery or fracture with one after leg trauma. Four patients had pre-existing symptoms of limb dysfunction but significantly worsened after immobilization or surgery. 23 percent of patients had immobilization or trauma of the affected limb. This might have implications for management of CBS, for avoiding injury, limiting immobilization and increasing movement in the affected limb. Copyright © 2012 Elsevier Ltd. All rights reserved.

  10. Advanced immobilization processes for fuel hulls and dissolver residues

    International Nuclear Information System (INIS)

    Hebel, W.; Boehme, G.; Findlay, J.R.; Sombret, C.

    1984-08-01

    Various research and development projects for the conditioning of cladding scraps and dissolver residues are pursued within the scope of the R and D programme on nuclear waste Management of the European Community. They include the characterization of the waste materials arising from industrial fuel reprocessing and the development of different waste immobilization techniques. These concern the embedment of scraps and residues into inert matrices like cement, metal alloys, compacted graphite and sintered ceramics as well as the treatment of the fuel hulls by melting or chemical conversion. The conditioned waste forms are tested as to their relevant properties for activity enclosure

  11. Advanced immobilization processes for fuel hulls and dissolver residues

    International Nuclear Information System (INIS)

    Hebel, W.; Boehme, G.; Findlay, J.R.; Sombert, C.

    1984-01-01

    Various research and development projects for the conditioning of cladding scraps and dissolver residues are pursued within the scope of the R and D programme on nuclear waste Management of the European Community. They include the characterization of the waste materials arising from industrial fuel reprocessing and the development of different waste immobilization techniques. These concern the embedment of the scraps and residues into inert matrices like cement, metal alloys, compacted graphite and sintered ceramics as well as the treatment of the fuel hulls by melting or chemical conversion. The conditioned waste forms are tested as to their relevant properties for activity enclosure

  12. Membranes suited for immobilizing biomolecules

    NARCIS (Netherlands)

    2009-01-01

    The present invention relates to flow-through membranes suitable for the immobilization of biomols., methods for the prepn. of such membranes and the use of such membranes for the immobilization of biomols. and subsequent detection of immobilized biomols. The invention concerns a flow-through

  13. Sustainable hybrid photocatalysts: titania immobilized on carbon materials derived from renewable and biodegradable resources

    Science.gov (United States)

    This review comprises the preparation, properties and heterogeneous photocatalytic applications of TiO2 immobilized on carbon materials derived from earth-abundant, renewable and biodegradable agricultural residues and sea food waste resources. The overview provides key scientifi...

  14. Natural and Synthetic Barriers to Immobilize Radionuclides

    International Nuclear Information System (INIS)

    Um, W.

    2011-01-01

    The experiments of weathering of glass waste form and the reacted sediments with simulated glass leachates show that radionuclide sequestration can be significantly enhanced by promoting the formation of secondary precipitates. In addition, synthetic phosphate-bearing nanoporous material exhibits high stability at temperature and has a very high K d value for U(VI) removal. Both natural and synthetic barrier materials can be used as additional efficient adsorbents for retarding transport of radionuclides for various contaminated waste streams and waste forms present at U. S. Department of Energy clean-up sites and the proposed geologic radioactive waste disposal facility. In the radioactive waste repository facility, natural or synthetic materials are planned to be used as a barrier material to immobilize and retard radionuclide release. The getter material can be used to selectively scavenge the radionuclide of interest from a liquid waste stream and subsequently incorporate the loaded getters in a cementitious or various monolithic waste forms. Also, the getter material is to reduce the release of radionuclides from monolithic waste forms. Also, the getter material is to reduce the release of radionuclides from monolithic waste forms. Also, the getter material is to reduce the release of radionuclides form monolithic waste forms by being emplaced as a backfill barrier material around the wastes or waste form to minimize the potential around the wastes or waste form to minimize the potential hazard of leached radioactive wastes. The barrier material should be highly efficient to sequester radionuclides and possess physical and chemical stability for long-term exposure to severe weathering conditions. Because potential leaching of radionuclides depends on various environmental and weathering conditions of the near-field repository, the barrier materials must be durable and not disintegrate under a range of moisture, temperature, pressure, radiation, Eh, ph. and

  15. Hanford Waste Vitrification Plant technology progress

    International Nuclear Information System (INIS)

    Wolfe, B.A.; Scott, J.L.; Allen, C.R.

    1989-10-01

    The Hanford Waste Vitrification Plant (HWVP) is currently being designed to safely process and temporarily store immobilized defense liquid high-level wastes from the Hanford Site. These wastes will be immobilized in a borosilicate glass waste form in the HWVP and stored onsite until a qualified geologic waste repository is ready for permanent disposal. Because of the diversity of wastes to be disposed of, specific technical issues are being addressed so that the plant can be designed and operated to produce a waste form that meets the requirements for permanent disposal in a geologic repository. This paper reports the progress to date in addressing these issues. 2 figs., 3 tabs

  16. HLW immobilization in glass: industrial operation and product quality

    International Nuclear Information System (INIS)

    Jacquet-Francillon, N.; Leroy, P.; Runge, S.

    1992-01-01

    This extended summary discusses the immobilization of high level wastes from the viewpoint of the quality of the final product, i.e. the HLW glass. The R and D studies comprise 3 steps: glass formulation, glass characterization and long term behaviour studies

  17. Evaluation of copper and lead immobilization in contaminated soil ...

    African Journals Online (AJOL)

    The effectiveness of natural clay, calcium phosphate, poultry manure and rice husks as cheap and ecologically non-invasive amendments for immobilizing Cu and Pb in contaminated soil was assessed. A moderately contaminated soil was sampled from a cultivated field in the vicinity of an active waste dump, characterized ...

  18. Programa de Aceleração do Crescimento (PAC: um estudo descritivo sobre os desperdícios ativos e passivos = Growth Acceleration Program (PAC: a descriptive study of active and passive waste

    Directory of Open Access Journals (Sweden)

    Daniele Silva Rodrigues

    2018-03-01

    Full Text Available Em 2007, o governo brasileiro implementou o Programa de Aceleração do Crescimento (PAC, com o objetivo de incentivar o crescimento econômico do Brasil e investimentos nas áreas de saneamento, habitação, transporte, energia e recursos hídricos. Tal programa, em sua primeira etapa (2007-2010, contribuiu de modo determinante para o aumento da oferta de empregos e da geração de renda no Brasil. No entanto, de acordo com Instituto Trata Brasil, 86% das obras de esgotamento sanitário provenientes do PAC estavam paralisadas, atrasadas ou não iniciadas até dezembro de 2012. Ainda, em 2008 e 2009, por meio da edição especial do Programa de Fiscalização por Sorteios Públicos, a Controladoria Geral da União (CGU constatou diversas irregularidades em 110 municípios brasileiros contemplados com recursos do PAC. Nesse contexto, o objetivo do estudo foi analisar as irregularidades constatadas pela CGU, na aplicação dos recursos federais destinados às áreas de saneamento e habitação dos municípios brasileiros por meio da primeira etapa do PAC. A principal fonte de dados da pesquisa foram os relatórios dos auditores da CGU advindos do PAC e dados do Censo 2000, disponibilizados pelo IBGE. Técnicas de análise de conteúdo foram utilizadas para classificar as irregularidades constatadas pela CGU em desperdício ativo (corrupção e passivo (má gestão e estatísticas descritivas para analisar o perfil das irregularidades e dos municípios auditados. Observou-se que os desperdícios mais frequentes nos municípios auditados são os classificados como Superfaturamento e Licitação Fraudada (classificados como desperdício ativo e Má Administração e Licitação Irregular, na modalidade de desperdício passivo. In 2007, the Brazilian government implemented the Growth Acceleration of Program (PAC, with the objective of encouraging Brazil's economic growth, with investments in the areas of sanitation, housing, transportation, energy and

  19. Immobilization of Uranium Silicides in Sintered Glass

    International Nuclear Information System (INIS)

    Mateos, P.; Russo, D.O.; Heredia, A.D.; Sanfilippo, M.

    2003-01-01

    High activity nuclear spent fuels vitrification by fusion is a well known technology which has industrial scale in France, England, Japan, EEUU. Borosilicates glasses are used in this process.Sintered glasses are an alternative to the immobilization task in which there is also a wide experience around the world.The available technics are: cold pressing and sintering , hot-pressing and hot isostatic pressing.This work compares Borosilicates and Iron silicates sintered glasses behaviour when different ammounts of nuclear simulated waste is added

  20. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Dixon, R.S.

    1984-12-01

    The Canadian Nuclear Fuel Waste Management Program involves research into the storage and transportation of used nuclear fuel, immobilization of fuel waste, and deep geological disposal of the immobilized waste. The program is now in the fourth year of a ten-year generic research and development phase. The objective of this phase of the program is to assess the safety and environmental aspects of the deep underground disposal of immobilized fuel waste in plutonic rock. The objectives of the research for each component of the program and the progress made to the end of 1983 are described in this report

  1. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 2. Alternatives for waste treatment

    International Nuclear Information System (INIS)

    1976-05-01

    Volume II of the five-volume report is devoted to the description of alternatives for waste treatment. The discussion is presented under the following section titles: fuel reprocessing modifications; high-level liquid waste solidification; treatment and immobilization of chop-leach fuel bundle residues; treatment of noncombustible solid wastes; treatment of combustible wastes; treatment of non-high-level liquid wastes; recovery of transuranics from non-high-level wastes; immobilization of miscellaneous non-high-level wastes; volatile radioisotope recovery and off-gas treatment; immobilization of volatile radioisotopes; retired facilities (decontamination and decommissioning); and, modification and use of selected fuel reprocessing wastes

  2. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 2. Alternatives for waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Volume II of the five-volume report is devoted to the description of alternatives for waste treatment. The discussion is presented under the following section titles: fuel reprocessing modifications; high-level liquid waste solidification; treatment and immobilization of chop-leach fuel bundle residues; treatment of noncombustible solid wastes; treatment of combustible wastes; treatment of non-high-level liquid wastes; recovery of transuranics from non-high-level wastes; immobilization of miscellaneous non-high-level wastes; volatile radioisotope recovery and off-gas treatment; immobilization of volatile radioisotopes; retired facilities (decontamination and decommissioning); and, modification and use of selected fuel reprocessing wastes. (JGB)

  3. Nuclear waste management. Quarterly progress report, July-September 1980

    Energy Technology Data Exchange (ETDEWEB)

    Chikalla, T.D.

    1980-11-01

    Research is reported on: high-level waste immobilization, alternative waste forms, TRU waste immobilization and decontamination, krypton solidification, thermal outgassing, /sup 129/I fixation, unsaturated zone transport, well-logging instrumentation, waste management system and safety studies, effectiveness of geologic isolation systems, waste/rock interactions, engineered barriers, backfill material, spent fuel storage (criticality), barrier sealing and liners for U mill tailings, and revegetation of inactive U tailings sites. (DLC)

  4. Plutonium immobilization plant using ceramic in existing facilities at the Savannah River site

    International Nuclear Information System (INIS)

    DiSabatino, A.

    1998-01-01

    The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources, and through a ceramic immobilization process converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. This immobilization process is shown conceptually in Figure 1-1. The objective is to make an immobilized form, suitable for geologic disposal, in which the plutonium is as inherently unattractive and inaccessible as the plutonium in spent fuel from commercial reactors. The ceramic immobilization alternative presented in this report consists of first converting the surplus material to an oxide, followed by incorporating the plutonium oxide into a titanate-based ceramic material that is placed in metal cans

  5. Waste disposal options report. Volume 1

    International Nuclear Information System (INIS)

    Russell, N.E.; McDonald, T.G.; Banaee, J.; Barnes, C.M.; Fish, L.W.; Losinski, S.J.; Peterson, H.K.; Sterbentz, J.W.; Wenzel, D.R.

    1998-02-01

    This report summarizes the potential options for the processing and disposal of mixed waste generated by reprocessing spent nuclear fuel at the Idaho Chemical Processing Plant. It compares the proposed waste-immobilization processes, quantifies and characterizes the resulting waste forms, identifies potential disposal sites and their primary acceptance criteria, and addresses disposal issues for hazardous waste

  6. Remote material handling in the Plutonium Immobilization Project. Revision 1

    International Nuclear Information System (INIS)

    Brault, J.R.

    2000-01-01

    With the downsizing of the US and Russian nuclear stockpiles, large quantities of weapons-usable plutonium in the US are being declared excess and will be disposed of by the Department of Energy Fissile Materials Disposition Program. To implement this program, DOE has selected the Savannah River Site (SRS) for the construction and operation of three new facilities: pit disassembly and conversion; mixed oxide fuel fabrication; and plutonium immobilization. The Plutonium Immobilization Project (PIP) will immobilize a portion of the excess plutonium in a hybrid ceramic and glass form containing high level waste for eventual disposal in a geologic repository. The PIP is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. Processing technology for the PIP is being developed jointly by the Lawrence Livermore National Laboratory and Westinghouse Savannah River Company. This paper will discuss development of the automated unpacking and sorting operations in the conversion area, and the automated puck and tray handling operations in the first stage immobilization area. Due to the high radiation levels and toxicity of the materials to be disposed of, the PIP will utilize automated equipment in a contained (glovebox) facility. Most operations involving plutonium-bearing materials will be performed remotely, separating personnel from the radiation source. Source term materials will be removed from the operations during maintenance. Maintenance will then be performed hands on within the containment using glove ports

  7. Imobilização prolongada e remobilização da articulação fêmoro-tíbio-patelar de ratos: estudo clínico e microscópico Experimental immobilization and remobilization rat knee joints: clinical and microscopic study

    Directory of Open Access Journals (Sweden)

    R.J. Del Carlo

    2007-04-01

    Full Text Available Trinta e quatro ratos foram alocados em quatro grupos experimentais: sem imobilização (G1, com imobilização do joelho direito por 45 dias (G2, com imobilização e remobilização com atividade livre por cinco semanas (G3, imobilização e remobilização com atividade livre e natação por cinco semanas (G4. A imobilização interferiu negativamente na marcha e amplitude articular e o G4 apresentou melhor evolução na marcha nos cinco primeiros dias, em relação ao G3. Após esse período, a evolução foi similar. Os componentes do G2 apresentaram rigidez articular, não observada em G3 e G4. Histologicamente, a imobilização promoveu aumento da espessura da cápsula articular, evidenciada pela presença do tecido conjuntivo fibroso que substituiu o tecido adiposo no G2, mas em menor proporção em G3 e G4. A imobilização determinou perda de proteoglicanos da matriz cartilaginosa, aumento do número de condrócitos, dispostos de forma irregular, aumento da espessura da cartilagem calcificada, irregularidade da superfície articular, proliferação de tecido conjuntivo no espaço intra-articular e aumento da espessura do osso subcondral. O G3 apresentou maior número de alterações na cartilagem e osso subcondral, quando comparado com G4. A imobilização degenerou as células sinoviais, indicando diminuição da produção de fluido sinovial e do suprimento nutricional à cartilagem. Tanto a atividade livre quanto sua associação com a natação favoreceram o retorno das condições biomecânicas e da cápsula articular, anteriores à imobilização.Thirty-four rats were randomly allocated into one of four experimental groups: without immobilization (G1, immobilization of the right knee joint for 45 days (G2, immobilization and remobilization with free activity for 5 weeks (G3, and immobilization and remobilization with free activity and swimming program for 5 weeks (G4. The immobilization was prejudicial to march and flexibility

  8. Immobilization of enzymes by radiation

    International Nuclear Information System (INIS)

    Kaetsu, I.; Kumakura, M.; Yoshida, M.; Asano, M.; Himei, M.; Tamura, M.; Hayashi, K.

    1979-01-01

    Immobilization of various enzymes was performed by radiation-induced polymerization of glass-forming monomers at low temperatures. Alpha-amylase and glucoamylase were effectively immobilized in hydrophilic polymer carrier such as poly(2-hydroxyethyl methacrylate) and also in rather hydrophobic carrier such as poly(tetraethylene-glycol diacrylate). Immobilized human hemoglobin underwent the reversible oxygenation concomitantly with change of oxygen concentration outside of the matrices. (author)

  9. Effects of immobilization on spermiogenesis

    Science.gov (United States)

    Meitner, E. R.

    1980-01-01

    The influence of immobilization stress on spermiogenesis in rats was investigated. After 96 hour immobilization, histological changes began to manifest themselves in the form of practically complete disappearance of cell population of the wall of seminiferous tubule as well as a markedly increased number of cells with pathologic mitoses. Enzymological investigations showed various changes of activity (of acid and alkaline phosphatase and nonspecific esterase) in the 24, 48, and 96 hour immobilization groups.

  10. Liquid secondary waste. Waste form formulation and qualification

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

  11. Plutonium Immobilization Program cold pour tests

    International Nuclear Information System (INIS)

    Hovis, G.L.; Stokes, M.W.; Smith, M.E.; Wong, J.W.

    1999-01-01

    The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site, Lawrence Livermore National Laboratory, Argonne National Laboratory, and Pacific Northwest National Laboratory to carry out the disposition of excess weapons-grade plutonium. This program uses the can-in-canister (CIC) approach. CIC involves encapsulating plutonium in ceramic forms (or pucks), placing the pucks in sealed stainless steel cans, placing the cans in long cylindrical magazines, latching the magazines to racks inside Defense Waste Processing Facility (DWPF) canisters, and filling the DWPF canisters with high-level waste glass. This process puts the plutonium in a stable form and makes it attractive for reuse. At present, the DWPF pours glass into empty canisters. In the CIC approach, the addition of a stainless steel rack, magazines, cans, and ceramic pucks to the canisters introduces a new set of design and operational challenges: All of the hardware installed in the canisters must maintain structural integrity at elevated (molten-glass) temperatures. This suggests that a robust design is needed. However, the amount of material added to the DWPF canister must be minimized to prevent premature glass cooling and excessive voiding caused by a large internal thermal mass. High metal temperatures, minimizing thermal mass, and glass flow paths are examples of the types of technical considerations of the equipment design process. To determine the effectiveness of the design in terms of structural integrity and glass-flow characteristics, full-scale testing will be conducted. A cold (nonradioactive) pour test program is planned to assist in the development and verification of a baseline design for the immobilization canister to be used in the PIP process. The baseline design resulting from the cold pour test program and CIC equipment development program will provide input to Title 1 design for second-stage immobilization. The cold pour tests will be conducted in two

  12. Plutonium Immobilization Project - Robotic canister loading

    International Nuclear Information System (INIS)

    Hamilton, R.L.

    2000-01-01

    The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site (SRS), Lawrence Livermore National Laboratory (LLNL), Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL). When operational in 2008, the PIP will fulfill the nation's nonproliferation commitment by placing surplus weapons-grade plutonium in a permanently stable ceramic form and making it unattractive for reuse. Since there are significant radiation and security concerns, the program team is developing novel and unique technology to remotely perform plutonium immobilization tasks. The remote task covered in this paper employs a jointed arm robot to load seven 3.5 inch diameter, 135-pound cylinders (magazines) through the 4 inch diameter neck of a stainless steel canister. Working through the narrow canister neck, the robot secures the magazines into a specially designed rack pre-installed in the canister. To provide the deterrent effect, the canisters are filled with a mixture of high-level waste and glass at the Defense Waste Processing Facility (DWPF)

  13. Preliminary assessment of blending Hanford tank wastes

    International Nuclear Information System (INIS)

    Geeting, J.G.H.; Kurath, D.E.

    1993-03-01

    A parametric study of blending Hanford tank wastes identified possible benefits from blending wastes prior to immobilization as a high level or low level waste form. Track Radioactive Components data were used as the basis for the single-shell tank (SST) waste composition, while analytical data were used for the double-shell tank (DST) composition. Limiting components were determined using the existing feed criteria for the Hanford Waste Vitrification Plant (HWVP) and the Grout Treatment Facility (GTF). Results have shown that blending can significantly increase waste loading and that the baseline quantities of immobilized waste projected for the sludge-wash pretreatment case may have been drastically underestimated, because critical components were not considered. Alternatively, the results suggest further review of the grout feed specifications and the solubility of minor components in HWVP borosilicate glass. Future immobilized waste estimates might be decreased substantially upon a thorough review of the appropriate feed specifications

  14. Preliminary assessment of blending Hanford tank wastes

    Energy Technology Data Exchange (ETDEWEB)

    Geeting, J.G.H.; Kurath, D.E.

    1993-03-01

    A parametric study of blending Hanford tank wastes identified possible benefits from blending wastes prior to immobilization as a high level or low level waste form. Track Radioactive Components data were used as the basis for the single-shell tank (SST) waste composition, while analytical data were used for the double-shell tank (DST) composition. Limiting components were determined using the existing feed criteria for the Hanford Waste Vitrification Plant (HWVP) and the Grout Treatment Facility (GTF). Results have shown that blending can significantly increase waste loading and that the baseline quantities of immobilized waste projected for the sludge-wash pretreatment case may have been drastically underestimated, because critical components were not considered. Alternatively, the results suggest further review of the grout feed specifications and the solubility of minor components in HWVP borosilicate glass. Future immobilized waste estimates might be decreased substantially upon a thorough review of the appropriate feed specifications.

  15. A new insight into the immobilization mechanism of Zn on biochar: the role of anions dissolved from ash

    OpenAIRE

    Tingting Qian; Yujun Wang; Tingting Fan; Guodong Fang; Dongmei Zhou

    2016-01-01

    Biochar is considered to be a promising material for heavy metal immobilization in soil. However, the immobilization mechanisms of Zn2+ on biochars derived from many common waste biomasses are not completely understood. Herein, biochars (denoted as PN350, PN550, WS350, and WS550) derived from pine needle (PN) and wheat straw (WS) were prepared at two pyrolysis temperatures (350??C and 550??C). The immobilization behaviors and mechanisms of Zn2+ on these biochars were systematically investigat...

  16. The mechanisms of heavy metal immobilization by cementitious material treatments and thermal treatments: A review.

    Science.gov (United States)

    Guo, Bin; Liu, Bo; Yang, Jian; Zhang, Shengen

    2017-05-15

    Safe disposal of solid wastes containing heavy metals is a significant task for environment protection. Immobilization treatment is an effective technology to achieve this task. Cementitious material treatments and thermal treatments are two types of attractive immobilization treatments due to that the heavy metals could be encapsulated in their dense and durable wasteforms. This paper discusses the heavy metal immobilization mechanisms of these methods in detail. Physical encapsulation and chemical stabilization are two fundamental mechanisms that occur simultaneously during the immobilization processes. After immobilization treatments, the wasteforms build up a low permeable barrier for the contaminations. This reduces the exposed surface of wastes. Chemical stabilization occurs when the heavy metals transform into more stable and less soluble metal bearing phases. The heavy metal bearing phases in the wasteforms are also reviewed in this paper. If the heavy metals are incorporated into more stable and less soluble metal bearing phases, the potential hazards of heavy metals will be lower. Thus, converting heavy metals into more stable phases during immobilization processes should be a common way to enhance the immobilization effect of these immobilization methods. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. Immobilization of microbial cell and yeast cell and its application to biomass conversion using radiation techniques

    International Nuclear Information System (INIS)

    Kaetsu, Isao; Kumakura, Minoru; Fujimura, Takashi; Kasai, Noboru; Tamada, Masao

    1987-01-01

    The recent results of immobilization of cellulase-producing cells and ethanol-fermentation yeast by radiation were reported. The enzyme of cellulase produced by immobilized cells was used for saccharification of lignocellulosic wastes and immobilized yeast cells were used for fermentation reaction from glucose to ethanol. The wastes such as chaff and bagasse were treated by γ-ray or electron-beam irradiation in the presence of alkali and subsequent mechanical crushing, to form a fine powder less than 50 μm in diameter. On the other hand, Trichoderma reesei as a cellulase-producing microbial cell was immobilized on a fibrous carrier having a specific porous structure and cultured to produce cellulase. The enzymatic saccharification of the pretreated waste was carried out using the produced cellulase. The enhanced fermentation process to produce ethanol from glucose with the immobilized yeast by radiation was also studied. The ethanol productivity of immobilized growing yeast cells thus obtained was thirteen times that of free yeast cells in a 1:1 volume of liquid medium to immobilized yeast cells. (author)

  18. Immobilization of microbial cell and yeast cell and its application to biomass conversion using radiation techniques

    Science.gov (United States)

    Kaetsu, Isao; Kumakura, Minoru; Fujimura, Takashi; Kasai, Noboru; Tamada, Masao

    The recent results of immobilization of cellulase-producing cells and ethanol-fermentation yeast by radiation were reported. The enzyme of cellulase produced by immobilized cells was used for saccharification of lignocellulosic wastes and immobilized yeast cells were used for fermentation reaction from glucose to ethanol. The wastes such as chaff and bagasse were treated by γ-ray or electron-beam irradiation in the presence of alkali and subsequent mechanical crushing, to form a fine powder less than 50 μm in diameter. On the other hand, Trichoderma reesei as a cellulase-producing microbial cell was immobilized on a fibrous carrier having a specific porous structure and cultured to produce cellulase. The enzymatic saccharification of the pretreated waste was carried out using the produced cellulase. The enhanced fermentation process to produce ethanol from glucose with the immobilized yeast by radiation was also studied. The ethanol productivity of immobilized growing yeast cells thus obtained was thirteen times that of free yeast cells in a 1:1 volume of liquid medium to immobilized yeast cells.

  19. Biodegradation of chlorobenzene using immobilized crude extracts ...

    African Journals Online (AJOL)

    SERVER

    2007-10-04

    Oct 4, 2007 ... immobilized crude extracts were reused for all other experiments and found that immobilization .... India which are of analytical reagent grade. .... 9. 60. 3. 1. Figure 3. Degradation of chlorobenzene by immobilized crude.

  20. Supramolecular protein immobilization on lipid bilayers

    NARCIS (Netherlands)

    Bosmans, R.P.G.; Hendriksen, W.E.; Verheijden, Mark Lloyd; Eelkema, R.; Jonkheijm, Pascal; van Esch, J.H.; Brunsveld, Luc

    2015-01-01

    Protein immobilization on surfaces, and on lipid bilayers specifically, has great potential in biomolecular and biotechnological research. Of current special interest is the immobilization of proteins using supramolecular noncovalent interactions. This allows for a reversible immobilization and

  1. Iodine immobilization in apatites

    International Nuclear Information System (INIS)

    Audubert, F.; Lartigue, J.E.

    2000-01-01

    In the context of a scientific program on long-lived radionuclide conditioning, a matrix for iodine 129 immobilization has been studied. A lead vanado-phosphate apatite was prepared from the melt of lead vanado-phosphate Pb 3 (VO 4 ) 1.6 (PO 4 ) 0.4 and lead iodide PbI 2 in stoichiometric proportions by calcination at 700 deg. C during 3 hours. Natural sintering of this apatite is not possible because the product decomposition occurs at 400 deg. C. Reactive sintering is the solution. The principle depends on the coating of lead iodide with lead vanado-phosphate. Lead vanado-phosphate coating is used as iodo-apatite reactant and as dense covering to confine iodine during synthesis. So the best condition to immobilize iodine during iodo-apatite synthesis is a reactive sintering at 700 deg. C under 25 MPa. We obtained an iodo-apatite surrounded with dense lead vanadate. Leaching behaviour of the matrix synthesized by solid-solid reaction is under progress in order to determine chemical durability, basic mechanisms of the iodo-apatite alteration and kinetic rate law. Iodo-apatite dissolution rates were pH and temperature dependent. We obtained a rate of 2.5 10 -3 g.m -2 .d -1 at 90 deg. C in initially de-ionised water. (authors)

  2. Waste management

    International Nuclear Information System (INIS)

    Dworschak, H.; Mannone, F.; Rocco, P.

    1995-01-01

    The presence of tritium in tritium-burning devices to be built for large scale research on thermonuclear fusion poses many problems especially in terms of occupational and environmental safety. One of these problems derives from the production of tritiated wastes in gaseous, liquid and solid forms. All these wastes need to be adequately processed and conditioned to minimize tritium releases to an acceptably low occupational and environmental level and consequently to protect workers and the public against the risks of unacceptable doses from exposure to tritium. Since all experimental thermonuclear fusion devices of the Tokomak type to be built and operated in the near future as well as all experimental activities undertaken in tritium laboratories like ETHEL will generate tritiated wastes, current strategies and practices to be applied for the routine management of these wastes need to be defined. Adequate background information is provided through an exhaustive literature survey. In this frame alternative tritiated waste management options so far investigated or currently applied to this end in Europe, USA and Canada have been assessed. The relevance of tritium in waste containing gamma-emitters, originated by the neutron activation of structural materials is assessed in relation to potential final disposal options. Particular importance has been attached to the tritium retention efficiency achievable by the various waste immobilization options. 19 refs., 2 figs., 1 tab

  3. Immobilization routes - they're not standing still

    Energy Technology Data Exchange (ETDEWEB)

    Basta, N

    1982-04-19

    A review of the current stage of research into enzyme immobilization and the application of this technology in food processing and biomass-energy conversion is presented. The major success of the technology at present is the production of high-fructose corn syrup in the U.S. A commercial-scale plant to make sweeteners from cheese whey using immobilized lactase has come onstream in the U.K. Of two other processes reported, one uses immobilized bacteria to treat waste-water and produces pipeline-quality methane, the other holds promise for cutting the cost of corn-to-ethanol processing and enhancing the performance of cellulose-to-ethanol routes.

  4. DWPF upgrade, immobilization Programmatic Environmental Impact Statement input. Revision 1

    International Nuclear Information System (INIS)

    Sullivan, I.K.; Bignell, D.

    1994-01-01

    This Programmatic Environmental Impact Statement (PEIS) addresses the immobilization of plutonium by vitrification. Existing engineering documents, analyses, EIS, and technical publications were used and incorporated wherever possible to provide a timely response to this support effort. Although the vitrification technology is proven for the immobilization of high-level radioactive waste, more study and technical detail will be necessary to provide a comprehensive EIS that fully addresses all aspects of introduction of plutonium to the vitrification process. This document describes the concept(s) of plutonium processing as it relates to the upgrade of the DWPF and is therefore conceptual in nature. These concepts are based on technical data and experience at the Savannah River Site and will be detailed and finalized to support execution of this immobilization option

  5. Purification, immobilization, and characterization of nattokinase on PHB nanoparticles.

    Science.gov (United States)

    Deepak, Venkataraman; Pandian, Suresh babu Ram Kumar; Kalishwaralal, Kalimuthu; Gurunathan, Sangiliyandi

    2009-12-01

    In this study, nattokinase was purified from Bacillus subtilis using ion exchange chromatography and immobilized upon polyhydroxybutyrate (PHB) nanoparticles. A novel strain isolated from industrial dairy waste was found to synthesize polyhydroxyalkanoates (PHA) and the strain was identified as Brevibacterium casei SRKP2. PHA granules were extracted from 48 h culture and the FT-IR analysis characterized them as PHB, a natural biopolymer from B. casei. Nanoprecipitation by solvent displacement technique was used to synthesize PHB nanoparticles. PHB nanoparticles were characterized using transmission electron microscopy and particle size ranged from 100-125 nm. Immobilization of nattokinase upon PHB nanoparticles resulted in a 20% increase in the enzyme activity. Immobilization also contributed to the enhanced stability of the enzyme. Moreover, the activity was completely retained on storage at 4 degrees C for 25 days. The method has proven to be highly simple and can be implemented to other enzymes also.

  6. Strategy for Qualification of the Plutonium Immobilized Form

    International Nuclear Information System (INIS)

    Marra, J.C.; Marra, S.L.; Bibler, N.E.; Strachan, D.M.; Shaw, H.F.

    1998-01-01

    In order to dispose of a radioactive waste form in a federal high- level waste repository, a waste form qualification strategy and program must be developed. The waste form qualification program must include the acceptance specifications for the product, show how compliance with these specifications will be met, and then demonstrate compliance. An important element of this program is developing a measure to demonstrate product consistency. For the can- in-canister option, waste form qualification is needed not only for the Pu immobilized form but also for the high-level waste glass canister containing the Pu waste form. The latter will require close interaction and coordination with the U.S. Department of Energy - Office of Environmental Management (DOE-EM) who, through their contractor at the Savannah River Site (SRS), is the producer of the high-level waste form at the Defense Waste Processing Facility (DWPF). In this paper a waste form qualification strategy for the plutonium ceramic form is described that utilizes, as much as possible, the qualification strategy successfully used for vitrified high-level waste

  7. Current activities in DOE's commercial waste management program

    International Nuclear Information System (INIS)

    1978-03-01

    Deep geologic disposal of radioactive wastes is being studied. Packaging and storage of spent fuel are also considered. Potential sites at Hanford and NTS are discussed. Research on waste immobilization and supporting studies is reported

  8. Experimental study on cesium immobilization in struvite structures.

    Science.gov (United States)

    Wagh, Arun S; Sayenko, S Y; Shkuropatenko, V A; Tarasov, R V; Dykiy, M P; Svitlychniy, Y O; Virych, V D; Ulybkina, Е А

    2016-01-25

    Ceramicrete, a chemically bonded phosphate ceramic, was developed for nuclear waste immobilization and nuclear radiation shielding. Ceramicrete products are fabricated by an acid-base reaction between magnesium oxide and mono potassium phosphate that has a struvite-K mineral structure. In this study, we demonstrate that this crystalline structure is ideal for incorporating radioactive Cs into a Ceramicrete matrix. This is accomplished by partially replacing K by Cs in the struvite-K structure, thus forming struvite-(K, Cs) mineral. X-ray diffraction and thermo-gravimetric analyses are used to confirm such a replacement. The resulting product is non-leachable and stable at high temperatures, and hence it is an ideal matrix for immobilizing Cs found in high-activity nuclear waste streams. The product can also be used for immobilizing secondary waste streams generated during glass vitrification of spent fuel, or the method described in this article can be used as a pretreatment method during glass vitrification of high level radioactive waste streams. Furthermore, it suggests a method of producing safe commercial radioactive Cs sources. Copyright © 2015 Elsevier B.V. All rights reserved.

  9. Experimental study on cesium immobilization in struvite structures

    International Nuclear Information System (INIS)

    Wagh, Arun S.; Sayenko, S.Y.; Shkuropatenko, V.A.; Tarasov, R.V.; Dykiy, M.P.; Svitlychniy, Y.O.; Virych, V.D.; Ulybkina, E.A.

    2016-01-01

    Graphical abstract: X-ray diffraction patterns of Ceramicrete forms, green representing struvite-K, and red, struvite-(K,Cs) with 10 wt.% CsCl in it. Cs substitutes partially for K, which immobilizes Cs at room temperature by the acid–base reaction. - Highlights: • Struvite structure of Ceramicrete is an excellent host of radioactive cesium. • The volatility problem of cesium can be avoided by this method. • This method can be used to produce cesium waste forms in ambient conditions. • It can also be used to pretreat cesium in glass vitrification technology. • It also provides a method to produce safe sealed radioactive sources of cesium. - Abstract: Ceramicrete, a chemically bonded phosphate ceramic, was developed for nuclear waste immobilization and nuclear radiation shielding. Ceramicrete products are fabricated by an acid–base reaction between magnesium oxide and mono potassium phosphate that has a struvite-K mineral structure. In this study, we demonstrate that this crystalline structure is ideal for incorporating radioactive Cs into a Ceramicrete matrix. This is accomplished by partially replacing K by Cs in the struvite-K structure, thus forming struvite-(K, Cs) mineral. X-ray diffraction and thermo-gravimetric analyses are used to confirm such a replacement. The resulting product is non-leachable and stable at high temperatures, and hence it is an ideal matrix for immobilizing Cs found in high-activity nuclear waste streams. The product can also be used for immobilizing secondary waste streams generated during glass vitrification of spent fuel, or the method described in this article can be used as a pretreatment method during glass vitrification of high level radioactive waste streams. Furthermore, it suggests a method of producing safe commercial radioactive Cs sources.

  10. Experimental study on cesium immobilization in struvite structures

    Energy Technology Data Exchange (ETDEWEB)

    Wagh, Arun S., E-mail: asw@anl.gov [Environmental Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, IL 60439 (United States); Sayenko, S.Y.; Shkuropatenko, V.A.; Tarasov, R.V.; Dykiy, M.P.; Svitlychniy, Y.O.; Virych, V.D.; Ulybkina, E.A. [National Science Center, Kharkov Institute of Physics and Technology, Kharkov (Ukraine)

    2016-01-25

    Graphical abstract: X-ray diffraction patterns of Ceramicrete forms, green representing struvite-K, and red, struvite-(K,Cs) with 10 wt.% CsCl in it. Cs substitutes partially for K, which immobilizes Cs at room temperature by the acid–base reaction. - Highlights: • Struvite structure of Ceramicrete is an excellent host of radioactive cesium. • The volatility problem of cesium can be avoided by this method. • This method can be used to produce cesium waste forms in ambient conditions. • It can also be used to pretreat cesium in glass vitrification technology. • It also provides a method to produce safe sealed radioactive sources of cesium. - Abstract: Ceramicrete, a chemically bonded phosphate ceramic, was developed for nuclear waste immobilization and nuclear radiation shielding. Ceramicrete products are fabricated by an acid–base reaction between magnesium oxide and mono potassium phosphate that has a struvite-K mineral structure. In this study, we demonstrate that this crystalline structure is ideal for incorporating radioactive Cs into a Ceramicrete matrix. This is accomplished by partially replacing K by Cs in the struvite-K structure, thus forming struvite-(K, Cs) mineral. X-ray diffraction and thermo-gravimetric analyses are used to confirm such a replacement. The resulting product is non-leachable and stable at high temperatures, and hence it is an ideal matrix for immobilizing Cs found in high-activity nuclear waste streams. The product can also be used for immobilizing secondary waste streams generated during glass vitrification of spent fuel, or the method described in this article can be used as a pretreatment method during glass vitrification of high level radioactive waste streams. Furthermore, it suggests a method of producing safe commercial radioactive Cs sources.

  11. Assessing attitudes toward spinal immobilization.

    Science.gov (United States)

    Bouland, Andrew J; Jenkins, J Lee; Levy, Matthew J

    2013-10-01

    Prospective studies have improved knowledge of prehospital spinal immobilization. The opinion of Emergency Medical Services (EMS) providers regarding spinal immobilization is unknown, as is their knowledge of recent research advances. To examine the attitudes, knowledge, and comfort of prehospital and Emergency Department (ED) EMS providers regarding spinal immobilization performed under a non-selective protocol. An online survey was conducted from May to July of 2011. Participants were drawn from the Howard County Department of Fire and Rescue Services and the Howard County General Hospital ED. The survey included multiple choice questions and responses on a modified Likert scale. Correlation analysis and descriptive data were used to analyze results. Comfort using the Kendrick Extrication Device was low among ED providers. Experienced providers were more likely to indicate comfort using this device. Respondents often believed that spinal immobilization is appropriate in the management of penetrating trauma to the chest and abdomen. Reported use of padding decreased along with the frequency with which providers practice and encounter immobilized patients. Respondents often indicated that they perform spinal immobilization due solely to mechanism of injury. Providers who feel as if spinal immobilization is often performed unnecessarily were more likely to agree that immobilization causes an unnecessary delay in patient care. The results demonstrate the need for improved EMS education in the use of the Kendrick Extrication Device, backboard padding, and spinal immobilization in the management of penetrating trauma. The attitudes highlighted in this study are relevant to the implementation of a selective spinal immobilization protocol. Copyright © 2013 Elsevier Inc. All rights reserved.

  12. Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Larson, D.E.; Allen, C.R.; Kruger, O.L.; Weber, E.T.

    1991-10-01

    The Hanford Waste Vitrification Plant (HWVP) is being designed to immobilize pretreated Hanford high-level waste and transuranic waste in borosilicate glass contained in stainless steel canisters. Testing is being conducted in the HWVP Technology Development Project to ensure that adapted technologies are applicable to the candidate Hanford wastes and to generate information for waste form qualification. Empirical modeling is being conducted to define a glass composition range consistent with process and waste form qualification requirements. Laboratory studies are conducted to determine process stream properties, characterize the redox chemistry of the melter feed as a basis for controlling melt foaming and evaluate zeolite sorption materials for process waste treatment. Pilot-scale tests have been performed with simulated melter feed to access filtration for solids removal from process wastes, evaluate vitrification process performance and assess offgas equipment performance. Process equipment construction materials are being selected based on literature review, corrosion testing, and performance in pilot-scale testing. 3 figs., 6 tabs

  13. Immobilization as a route to surplus fissile materials disposition

    International Nuclear Information System (INIS)

    Gray, L.W.; Kan, T.

    1995-01-01

    In the aftermath of the Cold War, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long-term management options, DOE has undertaken a multifaceted study to select options for storage and disposition of plutonium (Pu) in keeping with the national policy that Pu must be subjected to the highest standards of safety, security, and accountability. One alternative being considered is immobilization. To arrive at a suitable immobilization form, the authors first reviewed published information on high-level waste (HLW) immobilization technologies in order to identify 72 possible Pu immobilization forms to be prescreened. Surviving forms were screened using multiattribute analysis to determine the most promising technologies. Promising immobilization families were further evaluated to identify chemical, engineering, environmental, safety, and health problems that remain to be solved prior to making technical decisions as to the viability of using the form for long-term disposition of plutonium. All data, analyses, and reports are being provided to the DOE Fissile Materials Disposition Project Office to support the Record of Decision that is anticipated in the fourth quarter of FY96

  14. Immobilization of spent Bentonite by using cement matrix

    International Nuclear Information System (INIS)

    Isman MT; Endro-Kismolo

    1996-01-01

    Investigation of spent bentonite immobilization by using cement was done. The purpose of the investigation was to know the performance of cement in binding bentonite waste. The investigation was done by adding cement, water, and bentonite waste into a container and string until the mixture became homogenous. The mixture was put into a polyethylene tube (3.5 cm in diameter and 4 cm high) and it was cured up to 28 days. The specific weight of the monolith block was then calculated, and the compressive strength and the leaching rate in ground water and sea water was tested. The mass ratio of water to cement was 0.4. The variable investigated was the mass ratio of bentonite to cement. The immobilized bentonite waste was natural bentonite waste and activated bentonite waste. The result of the investigation showed that cement was good for binding bentonite waste. The maximum binding mass ratio of bentonite to cement was 0.4. In this condition the specific weight of the monolith block was 2.177 gram/cm 3 , its compressive strength was 22.6 N/mm 2 , and the leaching rate for 90 days in ground water and sea water was 5.7 x 10 -4 gram cm -2 day -1

  15. Nano-Immobilized Biocatalysts for Biodiesel Production from Renewable and Sustainable Resources

    Directory of Open Access Journals (Sweden)

    Keon Hee Kim

    2018-02-01

    Full Text Available The cost of biodiesel production relies on feedstock cost. Edible oil is unfavorable as a biodiesel feedstock because of its expensive price. Thus, non-edible crop oil, waste oil, and microalgae oil have been considered as alternative resources. Non-edible crop oil and waste cooking oil are more suitable for enzymatic transesterification because they include a large amount of free fatty acids. Recently, enzymes have been integrated with nanomaterials as immobilization carriers. Nanomaterials can increase biocatalytic efficiency. The development of a nano-immobilized enzyme is one of the key factors for cost-effective biodiesel production. This paper presents the technology development of nanomaterials, including nanoparticles (magnetic and non-magnetic, carbon nanotubes, and nanofibers, and their application to the nano-immobilization of biocatalysts. The current status of biodiesel production using a variety of nano-immobilized lipase is also discussed.

  16. Evaluation of The Use of Bentonite, Kaolin and Feldspar For Immobilizing The Uranium Radionuclide Slugdewaste

    International Nuclear Information System (INIS)

    Prayitno

    2006-01-01

    The experimental investigation on the mixture of bentonite, kaolin, feldspar, sludge waste and with the ratio of bentonite, kaolin, feldspar for evaluating its effectiveness has been done. Experimentally, this investigation is the preliminary study of the use of bentonite, kaolin, feldspar as a material for the immobilization of sludge waste containing material element especially uranium. This investigation was conducted by mixing waste (2,5; 5; 7.5; 10; 12.5 and 15 %) of total weight sludge waste and bentonite, kaolin, feldspar with of 800, 900 and 1000 o C temperature. The obtained the process condition in which the uranium fraction immobilized could be kept in the limit of safety standard for the sludge waste. Therefore, it could be concluded that the addition of hay ash as an additive in the formation of block monolith tend to minimize the leached sludge waste in the leaching media. (author)

  17. Application of spent fuel treatment technology to plutonium immobilization

    International Nuclear Information System (INIS)

    McPheeters, C.C.; Ackerman, J.P.; Gay, E.C., Johnson, G.K.

    1996-01-01

    The purpose of the electrometallurgical treatment technology being developed at Argonne National Laboratory (ANL) is to convert certain spent nuclear fuels into waste forms that are suitable for disposal in a geological repository for nuclear waste. The spent fuels of interest are those that cannot be safely stored for a long time in their current condition, and those that cannot be qualified for repository disposal. This paper explores the possibility of applying this electrometallurgical treatment technology to immobilization of surplus fissile materials, primarily plutonium. Immobilization of surplus fissile materials by electrometallurgical treatment could be done in the same facilities, at the same time. and in the same equipment as the proposed treatment of the present inventory of spent nuclear fuel. The cost and schedule savings of this simultaneous treatment scheme would be significant

  18. Conversion of radioactive ferrocyanide compounds to immobile glasses

    International Nuclear Information System (INIS)

    Schulz, W.W.; Dressen, A.L.

    1977-01-01

    Complex radioactive ferrocyanide compounds result from the scavenging of cesium from waste products produced in the chemical reprocessing of nuclear fuel. These ferrocyanides, in accordance with this process, are converted to an immobile glass, resistant to leaching by water, by fusion together with sodium carbonate and a mixture of (a) basalt and boron trioxide (B 2 O 3 ) or (b) silica (SiO 2 ) and lime (CaO). 7 claims

  19. Estudos de atmosfera

    Directory of Open Access Journals (Sweden)

    Carlos Eduardo Riccioppo

    2016-11-01

    Full Text Available Joker e Estudos de balística possuíam formas distintas dos cadernos ou livros quando foram exibidos pela primeira vez, na mostra Dual Overdrive. De algum modo, o flagrante das imagens dos cartazes da campanha política parisiense e daquelas cusparadas sobre o asfalto era preservado no modo como eram mostrados os trabalhos, que repunham não apenas a orientação espacial dos objetos fotografados, mas, igualmente, sua escala: Joker apresentava-se na parede, em dimensões relativamente próximas às dos cartazes lambe-lambe que retratavam; e Estudos de balística, no chão da galeria, também com ampliação suficiente para que se tivesse a impressão de que se tratava de imagens “em tamanho real”.

  20. Tank waste remediation system program plan

    International Nuclear Information System (INIS)

    Powell, R.W.

    1998-01-01

    This program plan establishes the framework for conduct of the Tank Waste Remediation System (TWRS) Project. The plan focuses on the TWRS Retrieval and Disposal Mission and is specifically intended to support the DOE mid-1998 Readiness to Proceed with Privatized Waste Treatment evaluation for establishing firm contracts for waste immobilization

  1. Tank waste remediation system program plan

    Energy Technology Data Exchange (ETDEWEB)

    Powell, R.W.

    1998-01-05

    This program plan establishes the framework for conduct of the Tank Waste Remediation System (TWRS) Project. The plan focuses on the TWRS Retrieval and Disposal Mission and is specifically intended to support the DOE mid-1998 Readiness to Proceed with Privatized Waste Treatment evaluation for establishing firm contracts for waste immobilization.

  2. Understanding radioactive waste

    International Nuclear Information System (INIS)

    Murray, R.L.

    1989-01-01

    This book discusses the sources and health effects of radioactive wastes. It reveals the techniques to concentrate and immobilize radioactivity and examines the merits of various disposal ideas. The book, which is designed for the lay reader, explains the basic science of atoms,nuclear particles,radioactivity, radiation and health effects

  3. Plutonium immobilization program - Cold pour Phase 1 test results

    International Nuclear Information System (INIS)

    Hamilton, L.

    2000-01-01

    The Plutonium Immobilization Project will disposition excess weapons grade plutonium. It uses the can-in-canister approach that involves placing plutonium-ceramic pucks in sealed cans that are then placed into Defense Waste Processing Facility canisters. These canisters are subsequently filled with high-level radioactive waste glass. This process puts the plutonium in a stable form and makes it unattractive for reuse. A cold (non-radioactive) glass pour program was performed to develop and verify the baseline design for the canister and internal hardware. This paper describes the Phase 1 scoping test results

  4. Plutonium Immobilization Program - Cold pour Phase 1 test results

    International Nuclear Information System (INIS)

    Hamilton, L.

    2000-01-01

    The Plutonium Immobilization Project will disposition excess weapons grade plutonium. It uses the can-in-canister approach that involves placing plutonium-ceramic pucks in sealed cans that are then placed into Defense Waste Processing Facility canisters. These canisters are subsequently filled with high-level radioactive waste glass. This process puts the plutonium in a stable form and makes it unattractive for reuse. A cold (non-radioactive) glass pour program was performed to develop and verify the baseline design for the canister and internal hardware. This paper describes the Phase 1 scoping test results

  5. Metastatic tumor of thoracic and lumbar spine: prospective study comparing the surgery and radiotherapy vs external immobilization with radiotherapy; Metastases do segmento toracico e lombar da coluna vertebral: estudo prospectivo comparativo entre o tratamento cirurgico e radioterapico com a imobilizacao externa e radioterapia

    Energy Technology Data Exchange (ETDEWEB)

    Falavigna, Asdrubal; Ioppi, Ana Elisa Empinotti; Grasselli, Juliana [Universidade de Caxias do Sul, RS (Brazil). Faculdade de Medicina]. E-mail: asdrubal@doctor.com; Righesso Neto, Orlando [Faculdade Federal de Ciencias Medicas da Santa Casa de Sao Paulo, SP (Brazil)

    2007-09-15

    Bone metastases at the thoracic and lumbar segment of the spine are usually presented with painful sensation and medullar compression. The treatment is based on the clinical and neurological conditions of the patient and the degree of tumor invasion. In the present study, 32 patients with spinal metastasis of thoracic and lumbar segment were prospectively analyzed. These patients were treated by decompression and internal stabilization followed by radiotherapy or irradiation with external immobilization. The election of the groups was in accordance with the tumor radiotherapy sensitivity, clinical conditions, spinal stability, medullar or nerve compression and patient's decision. The Frankel scale and pain visual test were applied at the moment of diagnosis and after 1 and 6 months. The surgical group had better results with preserving the ambulation longer and significant reduction of pain.(author)

  6. Ceramic Hosts for Fission Products Immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Peter C Kong

    2010-07-01

    Natural spinel, perovskite and zirconolite rank among the most leach resistant of mineral forms. They also have a strong affinity for a large number of other elements and including actinides. Specimens of natural perovskite and zirconolite were radioisotope dated and found to have survived at least 2 billion years of natural process while still remain their loading of uranium and thorium . Developers of the Synroc waste form recognized and exploited the capability of these minerals to securely immobilize TRU elements in high-level waste . However, the Synroc process requires a relatively uniform input and hot pressing equipment to produce the waste form. It is desirable to develop alternative approaches to fabricate these durable waste forms to immobilize the radioactive elements. One approach is using a high temperature process to synthesize these mineral host phases to incorporate the fission products in their crystalline structures. These mineral assemblages with immobilized fission products are then isolated in a durable high temperature glass for periods measured on a geologic time scale. This is a long term research concept and will begin with the laboratory synthesis of the pure spinel (MgAl2O4), perovskite (CaTiO3) and zirconolite (CaZrTi2O7) from their constituent oxides. High temperature furnace and/or thermal plasma will be used for the synthesis of these ceramic host phases. Nonradioactive strontium oxide will be doped into these ceramic phases to investigate the development of substitutional phases such as Mg1-xSrxAl2O4, Ca1-xSrxTiO3 and Ca1-xSrxZrTi2O7. X-ray diffraction will be used to establish the crystalline structures of the pure ceramic hosts and the substitution phases. Scanning electron microscopy and energy dispersive X-ray analysis (SEM-EDX) will be performed for product morphology and fission product surrogates distribution in the crystalline hosts. The range of strontium doping is planned to reach the full substitution of the divalent

  7. EIS Data Call Report: Plutonium immobilization plant using ceramic in new facilities at the Savannah River Site

    International Nuclear Information System (INIS)

    DiSabatino, A.

    1998-01-01

    The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. This immobilization process is shown conceptually in Figure 1-1. The objective is to make an immobilized form, suitable for geologic disposal, in which the plutonium is as inherently unattractive and inaccessible as the plutonium in spent fuel from commercial reactors. The ceramic immobilization alternative presented in this report consists of first converting the surplus material to an oxide, followed by incorporating the plutonium oxide into a titanate-based ceramic material that is placed in metal cans

  8. ICPP radioactive liquid and calcine waste technologies evaluation final report and recommendation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    Using a formalized Systems Engineering approach, the Latched Idaho Technologies Company developed and evaluated numerous alternatives for treating, immobilizing, and disposing of radioactive liquid and calcine wastes at the Idaho Chemical Processing Plant. Based on technical analysis data as of March, 1995, it is recommended that the Department of Energy consider a phased processing approach -- utilizing Radionuclide Partitioning for radioactive liquid and calcine waste treatment, FUETAP Grout for low-activity waste immobilization, and Glass (Vitrification) for high-activity waste immobilization -- as the preferred treatment and immobilization alternative.

  9. Immobilization of transuranic sludge in glass-ceramic materials

    International Nuclear Information System (INIS)

    Welch, J.M.; Schuman, R.P.; Flinn, J.E.

    1982-03-01

    Studies were performed to determine the effectiveness of glass-ceramic waste forms, particularly iron-enriched basalt, for immobilizing transuranic waste sludges from the Rocky Flats plant operations. Two sludges were used in the study - one was nonradioactive and the other contained approx. 2200 dps/mg of 241 Am. The glass-ceramic waste forms were produced from laboratory-scale melting operations with subsequent controlled cooling. The waste forms were examined to assess the microstructures which resulted from systematically varied compositions and controlled cooling sequences. Leach tests in deionized water were performed on small monolithic specimens of the various glass-ceramic materials. The test results showed a rather strong temperature dependence for leach rates. Also, for some of these materials, marked differences in the 241 Am leaching behavior were seen in measurements obtained on acidified versus neutral aliquots of the spent leachates. 8 figures, 12 tables

  10. HANFORD MEDIUM & LOW CURIE WASTE PRETREATMENT PROJECT PHASE 1 LAB REPORT

    Energy Technology Data Exchange (ETDEWEB)

    HAMILTON, D.W.

    2006-01-30

    A fractional crystallization (FC) process is being developed to supplement tank waste pretreatment capabilities provided by the Waste Treatment and Immobilization Plant (WTP). FC can process many tank wastes, separating wastes into a low-activity fraction (LAW) and high-activity fraction (HLW). The low-activity fraction can be immobilized in a glass waste form by processing in the bulk vitrification (BV) system.

  11. Storage and disposal of radioactive waste as glass in canisters

    International Nuclear Information System (INIS)

    Mendel, J.E.

    1978-12-01

    A review of the use of waste glass for the immobilization of high-level radioactive waste glass is presented. Typical properties of the canisters used to contain the glass, and the waste glass, are described. Those properties are used to project the stability of canisterized waste glass through interim storage, transportation, and geologic disposal

  12. High-level waste-form-product performance evaluation

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.; Allender, J.S.; Stone, J.A.; Gordon, D.E.; Gould, T.H. Jr.; Westberry, C.F. III.

    1982-01-01

    Seven candidate waste forms were evaluated for immobilization and geologic disposal of high-level radioactive wastes. The waste forms were compared on the basis of leach resistance, mechanical stability, and waste loading. All forms performed well at leaching temperatures of 40, 90, and 150 0 C. Ceramic forms ranked highest, followed by glasses, a metal matrix form, and concrete. 11 tables

  13. Estudos de atmosfera

    OpenAIRE

    Carlos Eduardo Riccioppo

    2016-01-01

    Joker e Estudos de balística possuíam formas distintas dos cadernos ou livros quando foram exibidos pela primeira vez, na mostra Dual Overdrive. De algum modo, o flagrante das imagens dos cartazes da campanha política parisiense e daquelas cusparadas sobre o asfalto era preservado no modo como eram mostrados os trabalhos, que repunham não apenas a orientação espacial dos objetos fotografados, mas, igualmente, sua escala: Joker apresentava-se na parede, em dimensões relativamente próximas às d...

  14. um estudo de caso

    OpenAIRE

    Costa, Cátia Filipa Pereira da

    2011-01-01

    Dissertação apresentada à Universidade Fernando Pessoa como parte dos requisitos para obtenção do grau de Mestre em Psicologia Jurídica As situações de abuso sexual de crianças nas quais o perpetrador pertence ao sexo feminino obtiveram ao longo dos últimos anos um acrescido reconhecimento por parte da comunidade científica, evidenciado pelo significativo incremento das investigações no âmbito desta temática consistindo na sua maioria estudos de caso. Um conjunto de particularidades encont...

  15. Cadmium immobilization by hydroxyapatite

    Directory of Open Access Journals (Sweden)

    Smičiklas Ivana D.

    2003-01-01

    Full Text Available The contamination of air, soil and water by cadmium is a great environmental problem. If cadmium occurs in nature in ionic form, soluble in water, it easily enters into the food chain. Hydroxyapatite (HAP, Ca-o(POAe(OH2 is a sparingly soluble salt and an excellent matrix for the removal of heavy metals from solutions. Considerable research attention has been paid to the bond between Cc/2+ ions and synthetic hydroxyapatite of known composition. The sorption mechanism is complex. The dominant process is ion exchange, but surface adsorption, surface complexation and coprecipitation can also contribute to the overall mechanism. The sorption capacity depends on the characteristics of hydroxyapatite itself and on the experimental conditions. Under optimum conditions a maximum capacity of 0.8 mol Cd2+/mol HAP can be achieved. HAP is a potential sorbent for the remediation of contaminated water and soil, for industrial waste treatment, and it is also referenced as a material that can be used as a barrier around waste depositories.

  16. Technological and organizational aspects of radioactive waste management

    International Nuclear Information System (INIS)

    2005-01-01

    This document comprises collected lecture on radioactive waste management which were given by specialists of the Radioactive Waste Management Section of the IAEA, scientific-industrial enterprise 'Radon' (Moscow, RF) and A.A. Bochvar's GNTs RF VNIINM (Moscow, RF) on various courses, seminars and conferences. These lectures include the following topics: basic principles and national systems of radioactive waste management; radioactive waste sources and their classification; collection, sorting and initial characterization of radioactive wastes; choice of technologies of radioactive waste processing and minimization of wastes; processing and immobilization of organic radioactive wastes; thermal technologies of radioactive waste processing; immobilization of radioactive wastes in cements, asphalts, glass and polymers; management of worked out closed radioactive sources; storage of radioactive wastes; deactivation methods; quality control and assurance in radioactive waste management

  17. Immobilization of cellulase by radiation polymerization

    International Nuclear Information System (INIS)

    Kumakura, M.; Kaetsu, I.

    1983-01-01

    Immobilization of cellulase by radiation polymerization at low temperatures was studied. The enzymatic activity of immobilized cellulase pellets varied with the monomer, enzyme concentration, and the thickness of immobilized cellulase pellets. The optimum monomer concentration in the immobilization of cellulase was 30-50% at the pellet thickness of 1.0 mm, in which the enzymatic activity was 50%. The enzymatic activity of immobilized cellulase pellets was examined using various substrates such as cellobiose, carboxymethylcellulose, and paper pretreated by radiation. It was found that irradiated paper can be hydrolyzed by immobilized cellulase pellets. (author)

  18. Synthesis of glycinamides using protease immobilized magnetic nanoparticles

    Directory of Open Access Journals (Sweden)

    Abha Sahu

    2016-12-01

    Full Text Available In the present investigation, Bacillus subtilis was isolated from slaughterhouse waste and screened for the production of protease enzyme. The purified protease was successfully immobilized on magnetic nanoparticles (MNPs and used for the synthesis of series of glycinamides. The binding and thermal stability of protease on MNPs was confirmed by FTIR spectroscopy and TGA analysis. The surface morphology of MNPs before and after protease immobilization was carried out using SEM analysis. XRD pattern revealed no phase change in MNPs after enzyme immobilization. The processing parameters for glycinamides synthesis viz. temperature, pH, and time were optimized using Response Surface Methodology (RSM by using Design Expert (9.0.6.2. The maximum yield of various amides 2 butyramidoacetic acid (AMD-1,83.4%, 2-benzamidoacetic acid (AMD-2,80.5% and 2,2′((carboxymethyl amino-2-oxoethyl-2-hydroxysuccinylbis(azanediyldiacetic acid (AMD-3,80.8% formed was observed at pH-8, 50 °C and 30 min. The synthesized immobilized protease retained 70% of the initial activity even after 8 cycles of reuse.

  19. Tank waste concentration mechanism study

    International Nuclear Information System (INIS)

    Pan, L.C.; Johnson, L.J.

    1994-09-01

    This study determines whether the existing 242-A Evaporator should continue to be used to concentrate the Hanford Site radioactive liquid tank wastes or be replaced by an alternative waste concentration process. Using the same philosophy, the study also determines what the waste concentration mechanism should be for the future TWRS program. Excess water from liquid DST waste should be removed to reduce the volume of waste feed for pretreatment, immobilization, and to free up storage capacity in existing tanks to support interim stabilization of SSTS, terminal cleanout of excess facilities, and other site remediation activities

  20. River Protection Project Mission Analysis Waste Blending Study

    International Nuclear Information System (INIS)

    Shuford, D.H.; Stegen, G.

    2010-01-01

    Preliminary evaluation for blending Hanford site waste with the objective of minimizing the amount of high-level waste (HLW) glass volumes without major changes to the overall waste retrieval and processing sequences currently planned. The evaluation utilizes simplified spreadsheet models developed to allow screening type comparisons of blending options without the need to use the Hanford Tank Waste Operations Simulator (HTWOS) model. The blending scenarios evaluated are expected to increase tank farm operation costs due to increased waste transfers. Benefit would be derived from shorter operating time period for tank waste processing facilities, reduced onsite storage of immobilized HLW, and reduced offsite transportation and disposal costs for the immobilized HLW.