WorldWideScience

Sample records for waste immobilization principles

  1. 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

  2. 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

  3. 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.)

  4. 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

  5. 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

  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. 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

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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.)

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. 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

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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

  14. 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 ...

  15. 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.

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  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. 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. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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.'

  14. 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.

  15. 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

  16. 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.

  17. 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.'

  18. 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)

  19. 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

  20. 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.

  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. 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

  3. 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

  4. 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

  5. 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

  6. 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%.

  7. 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)

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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.

  7. 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.)

  8. 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.)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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)

  2. 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

  3. 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)

  4. Principles of development of the industry of technogenic waste processing

    Directory of Open Access Journals (Sweden)

    Maria A. Bayeva

    2014-01-01

    Full Text Available Objective to identify and substantiate the principles of development of the industry of technogenic waste processing. Methods systemic analysis and synthesis method of analogy. Results basing on the analysis of the Russian and foreign experience in the field of waste management and environmental protection the basic principles of development activities on technogenic waste processing are formulated the principle of legal regulation the principle of efficiency technologies the principle of ecological safety the principle of economic support. The importance of each principle is substantiated by the description of the situation in this area identifying the main problems and ways of their solution. Scientific novelty the fundamental principles of development of the industry of the industrial wastes processing are revealed the measures of state support are proposed. Practical value the presented theoretical conclusions and proposals are aimed primarily on theoretical and methodological substantiation and practical solutions to modern problems in the sphere of development of the industry of technogenic waste processing.

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. 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).

  10. 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

  11. 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.

  12. 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.

  13. 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)

  14. 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.

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. Principles and guidelines for radioactive waste disposal facilities

    International Nuclear Information System (INIS)

    1988-06-01

    Four basic principles relevant to radioactive waste disposal identified. These principles cover the justification of the activity giving rise to the waste, the consideration of risk to present and future generations, the minimization of the need for intervention in the future, and the financial obligations of the licensee. The use of risk limits as opposed to dose limits associated with disposal is discussed, as are the concepts of critical group, de minimis, and ALARA, in the context of a waste disposal facility. Guidance is given on the selection of the preferred waste disposal concept from among several alternatives, and for judging proposed design improvements to the chosen concept

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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

  13. The principles of radioactive waste management. A publication within the RADWASS programme

    International Nuclear Information System (INIS)

    1995-01-01

    This publication defines the objective of radioactive waste management and the associated set of internationally agreed principles. The Safety Fundamentals include the objective of radioactive waste management and fundamental principles of radioactive waste management. The fundamental principles fall into the following general subject areas: protection of human health, protection of the environment, protection beyond national borders, responsibility to future generations and implementation procedures. Each principle is stated, and supporting and explanatory information pertaining to the principle is provided. 1 fig

  14. 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

  15. 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.

  16. 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.)

  17. 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.

  18. 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

  19. 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.

  20. Principles and objective of radioactive waste management

    International Nuclear Information System (INIS)

    Warnecke, E.

    1995-01-01

    Radioactive waste is generated in various nuclear applications, for example, in the use of radionuclides in medicine, industry and research or in the nuclear fuel cycle. It must be managed in a safe way independent of its very different characteristics. Establishing the basic safety philosophy is an important contribution to promoting and developing international consensus in radioactive waste management. The principles of radioactive waste management were developed with supporting text to provide such a safety philosophy. They cover the protection of human health and the environment now and in the future within and beyond national borders, the legal framework, the generation and management of radioactive wastes, and the safety of facilities. Details of the legal framework are provided by defining the roles and responsibilities of the Member State, the regulatory body and the waste generators and operators of radioactive waste management facilities. These principles and the responsibilities in radioactive waste management are contained in two recently published top level documents of the Radioactive Waste Safety Standards (RADWASS) programme which is the IAEA's contribution to foster international consensus in radioactive waste management. As the two documents have to cover all aspects of radioactive waste management they have to be formulated in a generic way. Details will be provided in other, more specific documents of the RADWASS programme as outlined in the RADWASS publication plant. The RADWASS documents are published in the Agency's Safety Series, which provides recommendations to Member Sates. Using material from the top level RADWASS documents a convention on the safety of radioactive waste management is under development to provide internationally binding requirements for radioactive waste management. (author). 12 refs

  1. 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.

  2. Waste Water Disposal Design And Management V

    International Nuclear Information System (INIS)

    Yang, Sang Hyeon; Lee, Jung Su

    2004-04-01

    This book deals with waste water disposal, design and management, which includes biofilm process, double living things treatment and microscopic organism's immobilized processing. It gives descriptions of biofilm process like construction, definition and characteristic of construction of biofilm process, system construction of biofilm process, principle of biofilm process, application of biofilm process, the basic treatment of double living thing and characteristic of immobilized processing of microscopic organism.

  3. 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)

  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. 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)

  6. 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

  7. 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

  8. 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.

  9. Principles and practices in managing the wastes resulting from decommissioning

    International Nuclear Information System (INIS)

    Vladescu, Gabriela; Oprescu, Theodor; Niculae, Ortenzia; Stan, Camelia

    2004-01-01

    The main objective in the management of radioactive wastes is the population and environment protection now and for the future without burdening the next generation with tasks other than their own. Achieving this objective is feasible if one takes into account the general principles internationally adopted and also the practices referring to the radioactive wastes, which can be summarized as: avoiding, minimizing, recovering, recycling, and storing. Minimizing the amount of wastes already produced resides in freeing part of them from the nuclear control by means of a process coined as classification. To implement such a process one must have in mind the premises required by classification and freeing the radioactive wastes from the regulating control, based on the legislation regarding the radioactive waste management and the measuring techniques and the corresponding procedures, as well. The target of this work was elaborating a proposal concerning the kind of classifying the radioactive waste in order to take them out from the nuclear control complying at the same time with the principles of minimizing and re-using as much as possible. The chapter 2.1 presents the frame of policy and regulations governing the process of management radioactive wastes. Here a proposal of classification of radioactive wastes is advanced based on the Romanian excepting levels adopted also by other countries, interpretation of the natural background, and the constraints concerning the radioactive and dangerous wastes. The chapter 2.2 presents the general principles of classifying the radioactive materials, of diluting the non-homogeneous distribution in solid materials as well as of the principles implied in the process of taking out some radioactive materials from the reach of regulating nuclear control. The chapter 2.3 deals with application of the radioactive waste management principles to reach a classification that entails taking these waste out from the reach of nuclear control

  10. 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

  11. 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...

  12. 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.)

  13. 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

  14. Waste minimization fundamental principles used in radioactive waste management plan for decommissioning of a CANDU - 600 nuclear power plant

    International Nuclear Information System (INIS)

    Barariu, Gheorghe; Georgescu, Roxana Cristiana; Sociu, Florin

    2009-01-01

    The objectives of waste minimization are to limit the generation and spread of radioactive contamination and to reduce the amount of wastes for storage and disposal, thereby limiting any consequent environmental impact, as well as the total costs associated with contaminated material management. This objective will be achieved by: reviewing the sources and characteristics of radioactive materials arising from Decontamination and Decommissioning (D and D) activities; reviewing waste minimization principles and current practical applications, together with regulatory, technical, financial and political factors influencing waste minimization practices; and reviewing current trends in improving waste minimization practices during Decontamination and Decommissioning. The main elements of a waste minimization strategy can be grouped into four areas: source reduction, prevention of contamination spread, recycle and reuse, and waste management optimization. For sustaining this objective, the following principles and procedures of wastes management are taken into account: safety and environment protection principles; principles regarding the facility operation; quality assurance procedures; procedures for material classification and releasing. (authors)

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. 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

  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. 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.

  4. 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

  5. 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

  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. 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

  8. 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)

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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

  15. 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)

  16. 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

  17. Criteria and principles for environmental assessment of disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Hill, M.D.

    1989-01-01

    This paper describes the criteria which are used in judging whether methods for the disposal of radioactive wastes are acceptable, from a radiological protection point of view, and the principles used in assessing the radiological impact of waste disposal methods. Gaseous, liquid and solid wastes are considered, and the discussion is relevant to wastes arising from the nuclear power industry, and from medical practices, general industry and research. Throughout the paper, emphasis is given to the general criteria and principles recommended by international organizations rather than to the detailed legislative and regulatory requirements in particular countries

  18. NSC confirms principles for safety review on Radioactive Waste Burial Facilities

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    The Nuclear Safety Commission authorized the scope of Principles for Safety Examination on Radioactive Waste Burial Facilities as suitable, the draft report for which was established by the Special Committee on Safety Standards of Radioactive Waste (Chairman Prof. Masao Sago, Science University of Tokyo) and reported on March 10 to the NSC. The principles include the theory that the facility must be controlled step by step, corresponding to the amount of radioactivity over 300 to 400 years after the burial of low-level solid radioactive waste with site conditions safe even in the event of occurrence of a natural disaster. The principles will be used for administrative safety examination against the application of the business on low-level radioactive waste burial facility which Japan Nuclear Fuel Industries, Inc. is planning to install at Rokkashomura, Aomori Prefecture. (author)

  19. 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

  20. 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

  1. 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.

  2. Basic principles and criteria on radioactive waste disposal sites

    International Nuclear Information System (INIS)

    Dlouhy, Z.; Kropikova, S.

    1980-01-01

    The basic principles are stated of radiation protection of the workers at radioactive waste disposal facilities, which must be observed in the choice of radioactive waste disposal sites. The emergency programme, the operating regulations and the safety report are specified. Workplace safety regulations are cited. (author)

  3. 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

  4. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    1983-01-01

    This report deals with certain aspects of the management of one of the most important wastes, i.e. the handling and storage of conditioned (immobilized and packaged) high-level waste from the reprocessing of spent nuclear fuel and, although much of the material presented here is based on information concerning high-level waste from reprocessing LWR fuel, the principles, as well as many of the details involved, are applicable to all fuel types. The report provides illustrative background material on the arising and characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The report introduces the principles important in conditioned high-level waste storage and describes the types of equipment and facilities, used or studied, for handling and storage of such waste. Finally, it discusses the safety and economic aspects that are considered in the design and operation of handling and storage facilities

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. Application of exemption principles to low-level waste disposal and recycle of wastes from nuclear facilities

    International Nuclear Information System (INIS)

    Kennedy, W.E. Jr.; Hemming, C.R.; O'Donnell, F.R.; Linsley, G.S.

    1988-04-01

    The International Atomic Energy Agency (IAEA) and other international groups are considering exempting from regulatory control certain radiation sources and practices, initially under the general heading of de minimis. A significant fraction of the wastes from industry, research, medicine, and the nuclear fuel cycle are contaminated to such low levels that the associated risks to health are trivial. IAEA work has been conducted by Advisory Groups to establish principles for exemption, and to apply the principles to various areas of waste management. In the second area, the main objectives have been to illustrate a methodology for developing practical radiological criteria through the application of the IAEA preliminary exemption principles, to establish generic criteria, and to determine the practicability of the preliminary exemption principles. The method used relies on a modeling assessment of the potential radiation exposure pathways and scenarios for individuals and population groups following the unrestricted release of materials. This paper describes the IAEA's assessment methodology and presents the generic results expressed in terms of the limiting activity concentration in municipal waste and in low-activity materials for recycle and reuse. 2 refs., 2 tabs

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. 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)

  16. 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

  17. 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)

  18. 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

  19. 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.

  20. 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

  1. Applications of exemption principles to low-level waste disposal and recycle of wastes from nuclear facilities

    International Nuclear Information System (INIS)

    Kennedy, W.E.; Hemming, C.R.; O'Donnell, F.R.; Linsley, G.S.

    1988-01-01

    The IAEA and other international groups for the past several years have been investigating the possibility of exempting from regulatory control certain radiation sources and practices, initially under the general heading of de minimis. IAEA work has been conducted by Advisory Groups on two interrelated levels : to establish principles for exemption, and to apply the principles to various areas of waste management. This paper describes the IAEA's assessment methodology and presents the generic results expressed in terms of the limiting activity concentrations in municipal waste and in low-activity materials for recycle and reuse

  2. 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.

  3. 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

  4. 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.

  5. 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

  6. Biogas from organically high polluted industrial waste waters

    Energy Technology Data Exchange (ETDEWEB)

    Sixt, H

    1985-06-01

    Organically high polluted waste water sets special claims for an economical purification and the process treatment. Up to now these waste waters are being purified by anaerobic processes with simultaneous biogas generation. The fourstep anaerobic degradation is influenced by a lot of important parameters. Extensive researchers in the field of anaerobic microbiology has improved the knowledge of the fundamental principles. Parallel the reactor technology is developed worldwide. In general it seems that the fixed-film-reactor with immobilized bacteria has the best future to purify organically high polluted industrial waste water with short retention times under stable operation conditions.

  7. 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)

  8. 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.

  9. 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

  10. 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.

  11. 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

  12. 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)

  13. 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)

  14. Principles of geological substantiation for toxic waste disposal facilities sites selection

    International Nuclear Information System (INIS)

    Khrushchov, D. P.; Matorin, Eu. M.; Shekhunova, S. B.

    2002-01-01

    Industrial, domestic and military activities result in accumulation of toxic and hazardous waste. Disposal of these waste comprises two main approaches: technological processing (utilization and destruction) and landfill. According to concepts and programs of advanced countries technological solutions are preferable, but in fact over 70 % of waste are buried in storages, prevailingly of near surface type. The target of this paper is to present principles of geological substantiation of sites selection for toxic and hazardous waste isolation facilities location. (author)

  15. The radiation protection principles model as a tool in the e-waste procedures

    Energy Technology Data Exchange (ETDEWEB)

    Tsitomeneas, S. Th., E-mail: stsit@teipir.gr [Piraeus University of Applied Sciences, Aigaleo (Greece); Vourlias, K., E-mail: kvourlias@yahoo.gr [Aristotle University of Thessaloniki (Greece); Geronikolou, St. A., E-mail: sgeronik@bioacademy.gr [Biomedical Research Foundation Academy of Athens, Athens (Greece)

    2016-03-25

    The electrical and electronic waste (e-waste) management is a global environmental problem dominated by the precautionary principle application, resulted to preliminary and ambiguous potential adverse effects, of extensive scientific uncertainty. In order to overcome the detected stochastic effects confusions in this field, we propose the inclusion of the principles of justification-optimization-limitation and of prudent avoidance. This model is already, established in radiation protection, so that toxicity as a result of the e-waste management would decrease, whilst the precious metals would be saved. We, further, resolve the classification of rejected items as reusable or as waste, so that the procedure of dismantling and recycling becomes easier, and the collecting-transporting-placement at an e-waste landfill would be safer. In conclusion, our proposing pattern in the e-waste management enforces the sustainable reducing-reusing-recycling, saves time/money and advances safety by including more sources of e-waste (military, medical etc) that were excluded previously.

  16. The radiation protection principles model as a tool in the e-waste procedures

    International Nuclear Information System (INIS)

    Tsitomeneas, S. Th.; Vourlias, K.; Geronikolou, St. A.

    2016-01-01

    The electrical and electronic waste (e-waste) management is a global environmental problem dominated by the precautionary principle application, resulted to preliminary and ambiguous potential adverse effects, of extensive scientific uncertainty. In order to overcome the detected stochastic effects confusions in this field, we propose the inclusion of the principles of justification-optimization-limitation and of prudent avoidance. This model is already, established in radiation protection, so that toxicity as a result of the e-waste management would decrease, whilst the precious metals would be saved. We, further, resolve the classification of rejected items as reusable or as waste, so that the procedure of dismantling and recycling becomes easier, and the collecting-transporting-placement at an e-waste landfill would be safer. In conclusion, our proposing pattern in the e-waste management enforces the sustainable reducing-reusing-recycling, saves time/money and advances safety by including more sources of e-waste (military, medical etc) that were excluded previously.

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. 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)

  2. 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.

  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. 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.

  5. 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

  6. 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

  7. Waste Immobilisation Plant (WIP), Trombay

    International Nuclear Information System (INIS)

    Kaushik, C.P.; Agarwal, K.

    2017-01-01

    Waste Immobilization Plant (WIP), Trombay is designed and constructed for the management of radioactive liquid wastes generated during reprocessing of spent nuclear fuel from research reactors at Bhabha Atomic Research Centre. In common with such facilities elsewhere, the objective here is to manage the wastes in such a way as to protect human health and the environment and to limit any burden on future generations. The plant has several facilities for the handling and treatment of the three classes of waste, viz., high, intermediate and low level, a classification based on their radioactivity content. In keeping with the general objective of radioactive waste management, the focus is on concentration and confinement of radioactivity. Strict adherence to the universal principles of radiation protection during operation of the plant ensures that radiation exposure is always kept as low as reasonably achievable (ALARA) under the prescribed limits

  8. 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.

  9. 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.

  10. 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.

  11. The use of mineral-like matrices for hlw solidification and spent fuel immobilization

    International Nuclear Information System (INIS)

    Pokhitonov, J.A.; Starchenko, V.A.; Strelnikov, A.V.; Sorokin, V.T.; Shvedov, A.A.

    2000-01-01

    The conception of radioactive waste management is based upon the multi-barrier protection principle stating that the long-lived radionuclides safety isolation is ensured by a system of engineering and natural geological barriers. One of the effective ways of the long-lived radionuclides immobilization is the integration of these materials within a mineral-like matrice. This technique may be used both for isolation of separated groups of nuclides (Cs, Sr, TUE, TRE) and for immobilization of spent fuel which for some reason can't be processed at the radiochemical plant. In this paper two variants of flowsheets HLW management are discussed. The following ways of HLW reprocessing are considered: - The first cycle raffinate solidification (without partitioning); - The individual solidification of two separated radionuclide groups (Sr+Cs+FP fraction and TPE+TRE fraction). The calcination of some characteristics (annual and total amounts, specific activity, radiochemical composition and radiogenic heat) of HLW integrated within a mineral-like matrix are performed for both options. The matrix compositions may be also used for spent fuel immobilization by means of the hot isostatic pressing technique. (authors)

  12. 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

  13. 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.

  14. Radioactive waste from non-licensed activities - identification of waste, compilation of principles and guidance, and proposed system for final management

    International Nuclear Information System (INIS)

    Jones, C.; Pers, K.

    2001-07-01

    Presently national guidelines for the handling of radioactive waste from non-licensed activities are lacking in Sweden. Results and information presented in this report are intended to form a part of the basis for decisions on further work within the Swedish Radiation Protection Institute on regulations or other guidelines on final management and final disposal of this type of waste. An inventory of radioactive waste from non-licensed activities is presented in the report. In addition, existing rules and principles used in Sweden - and internationally - on the handling of radioactive and toxic waste and non-radioactive material are summarized. Based on these rules and principles a system is suggested for the final management of radioactive material from non-licensed activities. A model is shown for the estimation of dose as a consequence of leaching of radio-nuclides from different deposits. The model is applied on different types of waste, e.g. peat ashes, light concrete and low-level waste from a nuclear installation

  15. 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.

  16. 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

  17. 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.

  18. 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.

  19. 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

  20. 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)

  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 [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)

  2. 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)

  3. 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)

  4. Legal Analysis of the Korea Radioactive Waste Management Act in the aspect of IAEA Principles

    International Nuclear Information System (INIS)

    Lee, D. S.; Chung, W. S.; Yang, M. H.; Yun, S. W.; Lee, J. H.

    2009-01-01

    According to the Principles of Radioactive Waste Management, the IAEA SAFETY SERIES NO-111-F, IAEA declared 9 doctrines. The IAEA advised a country that operates nuclear power plant to adopt the principles. As a member of the IAEA, Korea has also discussed about a unified policy and enacting law for radioactive waste management to follow the doctrines. This study analyzed the recently enacted Korea Radioactive Waste Management Act and verified whether the Act successfully follows the doctrine or not

  5. Application of clearance principles to radioactive waste from the decommissioning of nuclear reactors

    International Nuclear Information System (INIS)

    Lin Xiaoling; Feng Dingsheng; Dong Yonghe

    2010-01-01

    The definition of clearance is introduced. The principles and dose criterion of clearance are also clarified. The main radionuclides in radioactivity waste and the radioactivity waste which can be cleared are investigated. The techniques for the measurement of radioactivity waste from the decommissioning of nuclear reactors are summarized. This paper provides the scientific criterion and methods for the management of radioactive waste, and lays the foundation for the treatment of radioactive waste from the decommissioning of nuclear reactor. (authors)

  6. 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.

  7. 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

  8. 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

  9. 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}).

  10. 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

  11. 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.

  12. 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)

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. 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)

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. 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

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Reducing waste and errors: piloting lean principles at Intermountain Healthcare.

    Science.gov (United States)

    Jimmerson, Cindy; Weber, Dorothy; Sobek, Durward K

    2005-05-01

    The Toyota Production System (TPS), based on industrial engineering principles and operational innovations, is used to achieve waste reduction and efficiency while increasing product quality. Several key tools and principles, adapted to health care, have proved effective in improving hospital operations. Value Stream Maps (VSMs), which represent the key people, material, and information flows required to deliver a product or service, distinguish between value-adding and non-value-adding steps. The one-page Problem-Solving A3 Report guides staff through a rigorous and systematic problem-solving process. PILOT PROJECT at INTERMOUNTAIN HEALTHCARE: In a pilot project, participants made many improvements, ranging from simple changes implemented immediately (for example, heart monitor paper not available when a patient presented with a dysrythmia) to larger projects involving patient or information flow issues across multiple departments. Most of the improvements required little or no investment and reduced significant amounts of wasted time for front-line workers. In one unit, turnaround time for pathologist reports from an anatomical pathology lab was reduced from five to two days. TPS principles and tools are applicable to an endless variety of processes and work settings in health care and can be used to address critical challenges such as medical errors, escalating costs, and staffing shortages.

  15. 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.

  16. 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

  17. 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

  18. 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.

  19. Report for slot cutter proof-of-principle test, Buried Waste Containment System project. Revision 1

    International Nuclear Information System (INIS)

    1998-01-01

    Several million cubic feet of hazardous and radioactive waste was buried in shallow pits and trenches within many US Department of Energy (US DOE) sites. The pits and trenches were constructed similarly to municipal landfills with both stacked and random dump waste forms such as barrels and boxes. Many of the hazardous materials in these waste sites are migrating into groundwater systems through plumes and leaching. On-site containment is one of the options being considered for prevention of waste migration. This report describes the results of a proof-of-principle test conducted to demonstrate technology for containing waste. This proof-of-principle test, conducted at the RAHCO International, Inc., facility in the summer of 1997, evaluated equipment techniques for cutting a horizontal slot beneath an existing waste site. The slot would theoretically be used by complementary equipment designed to place a cement barrier under the waste. The technology evaluated consisted of a slot cutting mechanism, muck handling system, thrust system, and instrumentation. Data were gathered and analyzed to evaluate the performance parameters

  20. Report for slot cutter proof-of-principle test, Buried Waste Containment System project. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-05-21

    Several million cubic feet of hazardous and radioactive waste was buried in shallow pits and trenches within many US Department of Energy (US DOE) sites. The pits and trenches were constructed similarly to municipal landfills with both stacked and random dump waste forms such as barrels and boxes. Many of the hazardous materials in these waste sites are migrating into groundwater systems through plumes and leaching. On-site containment is one of the options being considered for prevention of waste migration. This report describes the results of a proof-of-principle test conducted to demonstrate technology for containing waste. This proof-of-principle test, conducted at the RAHCO International, Inc., facility in the summer of 1997, evaluated equipment techniques for cutting a horizontal slot beneath an existing waste site. The slot would theoretically be used by complementary equipment designed to place a cement barrier under the waste. The technology evaluated consisted of a slot cutting mechanism, muck handling system, thrust system, and instrumentation. Data were gathered and analyzed to evaluate the performance parameters.

  1. 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)

  2. 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.

  3. 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

  4. The precautionary principle and high-level nuclear waste policy

    International Nuclear Information System (INIS)

    Frishman, S.

    1999-01-01

    The 'Precautionary Principle' has grown from the broadening observation that there is compelling evidence that damage to humans and the world-wide environment is of such a magnitude and seriousness that new principles for conducting human activities are necessary. One of the various statements of the Precautionary Principle is: when an activity raises threats of harm to human health or the environment, precautionary measures should be taken even if some cause and effect relationships are not fully established scientifically. The use of a precautionary principle was a significant recommendation emerging from the 1992 United Nations Conference on Environment and Development, held in Rio de Janeiro, Brazil, and it is gaining acceptance in discussions ranging from global warming to activities that affect the marine environment, and far beyond. In the US high-level nuclear waste policy, there is a growing trend on the part of geologic repository proponents and regulators to shift the required safety evaluation from a deterministic analysis of natural and engineered barriers and their interactions to risk assessments and total system waste containment and isolation performance assessment. This is largely a result of the realisation that scientific 'proof' of safety cannot be demonstrated to the level repository proponents have led the American public to expect. Therefore, they are now developing other methods in an attempt to effectively lower the repository safety expectations of the public. Implicit in this shift in demonstration of 'proof' is that levels of uncertainty far larger than those generally taken as scientifically acceptable must be accepted in repository safety, simply because greater certainty is either too costly, in time and money, or impossible to achieve at the potential Yucca Mountain repository site. In the context of the Precautionary Principle, the repository proponent must bear the burden of providing 'Acceptable' proof, established by an open

  5. 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

  6. International principles for exemption from regulatory control and their application to waste management

    International Nuclear Information System (INIS)

    Linsley, G.S.

    1989-01-01

    This paper describes the main features of the international consensus on principles for exempting radiation sources and practices from regulatory control reached at a meeting in Vienna in March 1988. Some remaining problems, both of a philosophical and technical nature especially related to the application of the principles to radioactive waste disposal, are discussed

  7. Review: Waste-Pretreatment Technologies for Remediation of Legacy Defense Nuclear Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Wilmarth, William R.; Lumetta, Gregg J.; Johnson, Michael E.; Poirier, Micheal R.; Thompson, Major C.; Suggs, Patricia C.; Machara, N.

    2011-01-13

    The U.S. Department of Energy (DOE) is responsible for retrieving, immobilizing, and disposing of radioactive waste that has been generated during the production of nuclear weapons in the United States. The vast bulk of this waste material is stored in underground tanks at the Savannah River Site in South Carolina and the Hanford Site in Washington State. The general strategy for treating the radioactive tank waste consists of first separating the waste into high-level and low-activity fractions. This initial partitioning of the waste is referred to as pretreatment. Following pretreatment, the high-level fraction will be immobilized in a glass form suitable for disposal in a geologic repository. The low-activity waste will be immobilized in a waste form suitable for disposal at the respective site. This paper provides a review of recent developments in the application of pretreatment technologies to the processing of the Hanford and Savannah River radioactive tank wastes. Included in the review are discussions of 1) solid/liquid separations methods, 2) cesium separation technologies, and 3) other separations critical to the success of the DOE tank waste remediation effort. Also included is a brief discussion of the different requirements and circumstances at the two DOE sites that have in some cases led to different choices in pretreatment technologies.

  8. 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

  9. 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.

  10. Nuclear waste. DOE's program to prepare high-level radioactive waste for final disposal

    International Nuclear Information System (INIS)

    Bannerman, Carl J.; Owens, Ronald M.; Dowd, Leonard L.; Herndobler, Christopher S.; Purvine, Nancy R.; Stenersen, Stanley G.

    1989-11-01

    In summary, as of December 1988, the four sites collectively stored about 95 million gallons of high-level waste in underground tanks and bins. Approximately 57 million gallons are stored at Hanford, 34 million gallons at Savannah River, 3 million gallons at INEL, and 6 million gallons at West Valley. The waste is in several forms, including liquid, sludge, and dry granular materials, that make it unsuitable for permanent storage in its current state at these locations. Leaks from the tanks, designed for temporary storage, can pose an environmental hazard to surrounding land and water for thousands of years. DOE expects that when its waste processes at Savannah River, West Valley, and Hanford become operational, the high-level radioactive waste stored at these sites will be blended with other materials to immobilize it by forming a glass-like substance. The glass form will minimize the risk of environmental damage and make the waste more acceptable for permanent disposal in a geologic repository. At INEL, DOE is still considering various other immobilization and permanent disposal approaches. In July 1989, DOE estimated that it would cost about $13 billion (in fiscal year 1988 dollars) to retrieve, process, immobilize, and store the high-level waste until it can be moved to a permanent disposal site: about $5.3 billion is expected to be spent at Savannah River, $0.9 billion at West Valley, $2.8 billion at Hanford, and $4.0 billion at INEL. DOE has started construction at Savannah River and West Valley for facilities that will be used to transform the waste into glass (a process known as vitrification). These sites have each encountered schedule delays, and one has encountered a significant cost increase over earlier estimates. More specifically, the Savannah River facility is scheduled to begin high-level waste vitrification in 1992; the West Valley project, based on a January 1989 estimate, is scheduled to begin high-level waste vitrification in 1996, about 8

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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...

  16. 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.

  17. Immobilization of Candida rugosa lipase by adsorption-crosslinking onto corn husk

    Science.gov (United States)

    Nuraliyah, A.; Wijanarko, A.; Hermansyah, H.

    2018-04-01

    Corn husk is one of the agricutural waste that has not been used optimally. corn husk waste allows to be used as immobilized support for biocatalyst because it is easy to obtain, available abundant, renewable and easy to decompose. This research was conducted in two phases, namely the adsorption of enzyme immobilization on the support, followed by cross- linking between the enzyme and support through the addition of glutaraldehyde. The optimum conditions for cross-linked adsorption immobilization using support of corn husk were achieved at concentrations of 0,75 mg / ml at 4 hour reaction time. The biggest unit activity value is obtained at 2,37 U / g support through 0.5% glutaraldehyde addition.

  18. 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

  19. 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)

  20. 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.

  1. 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)

  2. The principles of design of a shallow disposal site for low and intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    Holmes, R.E.

    1985-01-01

    This paper addresses the principles of design of a shallow disposal site for low and intermediate level radioactive wastes. The objective of the author is to review the need for shallow land disposal facilities in the UK and to propose design principles which will protect the public and operatives from excessive risk. It is not the intent of the author to present a detailed design of facility which will meet the design standards proposed although such a design is feasible and within the scope of currently available technology. The principles and standards proposed in this paper are not necessarily those of PPC Consultant Services Ltd. or NEI Waste Technologies Ltd. (author)

  3. Low-level tank waste simulant data base

    International Nuclear Information System (INIS)

    Lokken, R.O.

    1996-04-01

    The majority of defense wastes generated from reprocessing spent N- Reactor fuel at Hanford are stored in underground Double-shell Tanks (DST) and in older Single-Shell Tanks (SST) in the form of liquids, slurries, sludges, and salt cakes. The tank waste remediation System (TWRS) Program has the responsibility of safely managing and immobilizing these tank wastes for disposal. This report discusses three principle topics: the need for and basis for selecting target or reference LLW simulants, tanks waste analyses and simulants that have been defined, developed, and used for the GDP and activities in support of preparing and characterizing simulants for the current LLW vitrification project. The procedures and the data that were generated to characterized the LLW vitrification simulants were reported and are presented in this report. The final section of this report addresses the applicability of the data to the current program and presents recommendations for additional data needs including characterization and simulant compositional variability studies

  4. Basic safety principles of INSAG and their application in radioactive waste management

    International Nuclear Information System (INIS)

    Baer, A.J.

    2000-01-01

    The International Nuclear Safety Advisory Group (INSAG) has, in INSAG-11, attempted to show what safety principles are common to all applications of all sources of radiation. It has been considered that these general principles should apply to all industrial activities. A comparison of INSAG-11 with Article 11 of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (Joint Convention) shows that the management of radioactive waste is but a special case of industrial activity and follows the same safety rules. The importance of the Joint Convention comes, however, from the fact that it is a politically important document, requiring ratification by the parliaments of the contracting parties. The safe management of radioactive waste implies that five types of issue must be taken into consideration, not only technical and ethical ones, but also socio-political, economic and ecological ones. By comparison, sustainable development in its three dimensions (temporal, spatial and sectorial) has five components (ecology, economics, ethics, socio-politics and technology), just like the safe management of radioactive waste. The consequence of this is that if management is treated as a particular case of sustainable development, it will not be accepted by society. The conclusions are that technology alone can not ensure the safety of radioactive waste management and that society will always give priority to socio-political issues over technological ones. Furthermore, it is crucial that people involved in the management of radioactive waste learn to communicate better and to listen more attentively. Their efforts will only succeed when they incorporate all the components that determine the fabric of our society. (author)

  5. 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

  6. Hanford immobilized LAW product acceptance: Initial Tanks Focus Area testing data package

    Energy Technology Data Exchange (ETDEWEB)

    JD Vienna; A Jiricka; BP McGrail; BM Jorgensen; DE Smith; BR Allen; JC Marra; DK Peeler; KG Brown; IA Reamer; WL Ebert

    2000-03-08

    The Hanford Site's mission has been to produce nuclear materials for the US Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during plutonium 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 total volume of LAW requiring immobilization will include the LAW separated from the tank waste, as well as new wastes generated by the retrieval, pretreatment, and immobilization processes. Per the Tri-Party Agreement (1994), both the LAW and HLW will be vitrified. It has been estimated that vitrification of the LAW waste will result in over 500,000 metric tons or 200,000 m{sup 3} of immobilized LAW (ILAW) glass. The ILAW glass is to be disposed of onsite in a near-surface burial facility. It must be demonstrated that the disposal system will adequately retain the radionuclides and prevent contamination of the surrounding environment. This report describes a study of the impacts of systematic glass-composition variation on the responses from accelerated laboratory corrosion tests of representative LAW glasses. A combination of two tests, the product consistency test and vapor-hydration test, is being used to give indictations of the relative rate at which a glass could be expected to corrode in the burial scenario.

  7. Hanford immobilized LAW product acceptance: Initial Tanks Focus Area testing data package

    International Nuclear Information System (INIS)

    JD Vienna; A Jiricka; BP McGrail; BM Jorgensen; DE Smith; BR Allen; JC Marra; DK Peeler; KG Brown; IA Reamer; WL Ebert

    2000-01-01

    The Hanford Site's mission has been to produce nuclear materials for the US Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during plutonium 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 total volume of LAW requiring immobilization will include the LAW separated from the tank waste, as well as new wastes generated by the retrieval, pretreatment, and immobilization processes. Per the Tri-Party Agreement (1994), both the LAW and HLW will be vitrified. It has been estimated that vitrification of the LAW waste will result in over 500,000 metric tons or 200,000 m 3 of immobilized LAW (ILAW) glass. The ILAW glass is to be disposed of onsite in a near-surface burial facility. It must be demonstrated that the disposal system will adequately retain the radionuclides and prevent contamination of the surrounding environment. This report describes a study of the impacts of systematic glass-composition variation on the responses from accelerated laboratory corrosion tests of representative LAW glasses. A combination of two tests, the product consistency test and vapor-hydration test, is being used to give indictations of the relative rate at which a glass could be expected to corrode in the burial scenario

  8. Encapsulation of hazardous wastes into agglomerates

    International Nuclear Information System (INIS)

    Guloy, A.

    1992-01-01

    The objective of this study was to investigate the feasibility of using the cementitious properties and agglomeration characteristics of coal conversion byproducts to encapsulate and immobilize hazardous waste materials. The intention was to establish an economical way of co-utilization and co-disposal of wastes. In addition, it may aid in the eradication of air pollution problems associated with the fine-powdery nature of fly ash. Encapsulation into agglomerates is a novel approach of treating toxic waste. Although encapsulation itself is not a new concept, existing methods employ high-cost resins that render them economically unfeasible. In this investigation, the toxic waste was contained in a concrete-like matrix whereby fly ash and other cementitious waste materials were utilized. The method incorporates the principles of solidification, stabilization and agglomeration. Another aspect of the study is the evaluation of the agglomeration as possible lightweight aggregates. Since fly ash is commercially used as an aggregate, it would be interesting to study the effect of incorporating toxic wastes in the strength development of the granules. In the investigation, the fly ash self-cementation process was applied to electroplating sludges as the toxic waste. The process hoped to provide a basis for delisting of the waste as hazardous and, thereby greatly minimize the cost of its disposal. Owing to the stringent regulatory requirements for hauling and disposal of hazardous waste, the cost of disposal is significant. The current practice for disposal is solidifying the waste with portland cement and dumping the hardened material in the landfill where the cost varies between $700--950/ton. Partially replacing portland cement with fly ash in concrete has proven beneficial, therefore applying the same principles in the treatment of toxic waste looked very promising

  9. 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

  10. 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.

  11. 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)

  12. 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.

  13. 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.'

  14. 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.'

  15. 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.

  16. 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

  17. 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.

  18. Lean principles applied to software development – avoiding waste

    Directory of Open Access Journals (Sweden)

    Ionel NAFTANAILA

    2009-12-01

    Full Text Available Under the current economic conditions many organizations strive to continue the trend towards adopting better software development processes, in order to take advantage of the numerous benefits that these can offer. Those benefits include quicker return on investment, better software quality, and higher customer satisfaction. To date, however, there is little body of research that can guide organizations in adopting modern software development practices, especially when it comes to Lean thinking and principles. To address this situation, the current paper identifies and structures the main wastes (or muda in Lean terms in software development as described by Lean principles, in an attempt to bring into researchers’ and practitioners’ attention Lean Software Development, a modern development methodology based on well-established practices such as Lean Manufacturing or Toyota Production System.

  19. 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...

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor

    International Nuclear Information System (INIS)

    Garcia Martinez, H.

    1999-01-01

    In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 L ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

  5. The application of assessment principles to an operational low level waste disposal site in England

    International Nuclear Information System (INIS)

    McHugh, J.O.; Newstead, S.; Weedon, C.J.

    1988-01-01

    This paper reviews the current assessment principles utilized in England and discusses their application to the Drigg low-level Radioactive Waste Disposal Site. The Drigg Site was established in 1959 and the assessment principles were published in 1985; therefore, although the Drigg Site has operated successfully, the application of the assessment principles has caused changes in operations and the establishment of further site research by the Department of the Environment

  6. Safety principles and technical criteria for the underground disposal of high level radioactive wastes

    International Nuclear Information System (INIS)

    1989-01-01

    The main objective of this book is to set out an internationally agreed set of principles and criteria for the design of deep underground repositories for the disposal of high level radioactive wastes. This book is concerned with the post-closure period. Consideration of the operational requirements which must be met when wastes are being handled, stored and emplaced are not therefore included

  7. 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)

  8. 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.

  9. Applying the principles of thermoeconomics to the organic Rankine Cycle for low temperature waste heat recovery

    International Nuclear Information System (INIS)

    Xiao, F.; Lilun, Q.; Changsun, S.

    1989-01-01

    In this paper, thermoeconomic principle is used to study the selection of working fluids and the option of the cycle parameters in the organic Rankine cycle of low temperature waste heat recovery. The parameter ξ, the product of the ratio of waste heat recovery and real cycle thermal efficiency, is suggested as a unified thermodynamic criterion for the selection of the working fluids. The mathematical expressions are developed to determine the optimal boiling temperature and the optimal pin point temperature difference in the heat recovery exchanger by way of thermoeconomic principle

  10. 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.

  11. 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.

  12. Material chemistry challenges in vitrification of high level radioactive waste

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2008-01-01

    Full text: Nuclear technology with an affective environmental management plan and focused attention on safety measures is a much cleaner source of electricity generation as compared to other sources. With this perspective, India has undertaken nuclear energy program to share substantial part of future need of power. Safe containment and isolation of nuclear waste from human environment is an indispensable part of this programme. Majority of radioactivity in the entire nuclear fuel cycle is high level radioactive liquid waste (HLW), which is getting generated during reprocessing of spent nuclear fuels. A three stage strategy for management of HLW has been adopted in India. This involves (i) immobilization of waste oxides in stable and inert solid matrices, (ii) interim retrievable storage of the conditioned waste product under continuous cooling and (iii) disposal in deep geological formation. Borosilicate glass matrix has been adopted in India for immobilization of HLW. Material issue are very important during the entire process of waste immobilization. Performance of the materials used in nuclear waste management determines its safety/hazards. Material chemistry therefore has a significant bearing on immobilization science and its technological development for management of HLW. The choice of suitable waste form to deploy for nuclear waste immobilization is difficult decision and the durability of the conditioned product is not the sole criterion. In any immobilization process, where radioactive materials are involved, the process and operational conditions play an important role in final selection of a suitable glass formulation. In remotely operated vitrification process, study of chemistry of materials like glass, melter, materials of construction of other equipment under high temperature and hostile corrosive condition assume significance for safe and un-interrupted vitrification of radioactive to ensure its isolation waste from human environment. The present

  13. 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)

  14. 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)

  15. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-12

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

  17. 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

  18. 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)

  19. A process for treatment of mixed waste containing chemical plating wastes

    International Nuclear Information System (INIS)

    Anast, K.R.; Dziewinski, J.; Lussiez, G.

    1995-01-01

    The Waste Treatment and Minimization Group at Los Alamos National Laboratory has designed and will be constructing a transportable treatment system to treat low-level radioactive mixed waste generated during plating operations. The chemical and plating waste treatment system is composed of two modules with six submodules, which can be trucked to user sites to treat a wide variety of aqueous waste solutions. The process is designed to remove the hazardous components from the waste stream, generating chemically benign, disposable liquids and solids with low level radioactivity. The chemical and plating waste treatment system is designed as a multifunctional process capable of treating several different types of wastes. At this time, the unit has been the designated treatment process for these wastes: Destruction of free cyanide and metal-cyanide complexes from spent plating solutions; destruction of ammonia in solution from spent plating solutions; reduction of Cr VI to Cr III from spent plating solutions, precipitation, solids separation, and immobilization; heavy metal precipitation from spent plating solutions, solids separation, and immobilization, and acid or base neutralization from unspecified solutions

  20. 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.

  1. 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.

  2. Disposal of radwastes and recycling of wastes and structural materials -fundamental principles, concepts, results

    International Nuclear Information System (INIS)

    Schaller, G.; Arens, G.; Brennecke, P.; Goertz, R.; Poschner, J.; Thieme, M.

    1997-01-01

    This report describes the German concept for the disposal of radioactive waste, and the re-use or recycling of contaminated materials. All radioactive waste can be disposed of in deep geological formations (practised at ERAM disposal site, planned for Konrad disposal site). Radioactively contaminated material below clearance levels can proceed for disposal at waste disposal sites and incineration plants, or for re-use and recycling, especially where the material consists of contaminated steel and of buildings. The basic principles (dose limits and model structures for deriving recommendations), reference values, or limits are described. The latest concepts are described in greater detail. Waste management in Germany is compared with international concepts. (orig.) [de

  3. 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

  4. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  5. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a join ted arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  6. The main ecological principles of ensuring safety of man and biosphere in the handling of radioactive wastes

    International Nuclear Information System (INIS)

    Kryshev, I.I.; Sazykina, T.G.

    1999-01-01

    This paper provides an assessment of ecological safety in the handling of radioactive wastes in the territory of Russia. The following problems are considered: the main sources of radioactive wastes and spent nuclear fuel; assessments of collective dose from the enterprises of the nuclear fuel cycle in Russia; and principles and criteria for ensuring ecological safety when handling radioactive wastes

  7. 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

  8. Life cycle costing of waste management systems: Overview, calculation principles and case studies

    DEFF Research Database (Denmark)

    Martinez Sanchez, Veronica; Kromann, Mikkel A.; Astrup, Thomas Fruergaard

    2015-01-01

    This paper provides a detailed and comprehensive cost model for the economic assessment of solid waste management systems. The model was based on the principles of Life Cycle Costing (LCC) and followed a bottom-up calculation approach providing detailed cost items for all key technologies within...... regarding the cost assessment of waste management, namely system boundary equivalency, accounting for temporally distributed emissions and impacts, inclusions of transfers, the internalisation of environmental impacts and the coverage of shadow prices, and there was also significant confusion regarding...

  9. 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

  10. Corrosion of steel drums containing immobilized ion exchange-resins and incineration ashes

    International Nuclear Information System (INIS)

    Marotta, F.; Schulz Rodriguez, F.M.; Farina, Silvia B.; Duffo, Gustavo S.

    2009-01-01

    The Argentine Atomic Energy Commission (CNEA) is responsible for developing the management nuclear waste disposal programme. This programme contemplates the design and construction of a facility for the final disposal of intermediate-level radioactive wastes. The proposed model is a near-surface monolithic repository similar to those in operation in El Cabril, Spain. The design of this type of repository is based on the use of multiple, independent and redundant barriers. The intermediate radioactive waste consists mostly in spent ionic exchange resins and filters from the nuclear power plants, research reactors and radioisotopes production facilities. The spent resins, as well as the incineration ashes, have to be immobilized before being stored to improve leach resistance of waste matrix and to maintain mechanical stability for safety requirements. Generally, cementation processes have been used as immobilization techniques for economical reasons as well as for being a simple operation. The immobilized resins and incineration ashes are thus contained in steel drums that, in turn, can undergo corrosion depending on the ionic content of the matrix. This work is a part of a systematic study of the corrosion susceptibility of steel drums in contact with immobilized cemented exchange-resins with different types and contents of aggressive species and incineration ashes. To this purpose, a special type of specimen was manufactured to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix are being monitored along time. The aggressive species studied were chloride ions (the main ionic species present in nature) and sulphate ions (produced during the radiolysis process of the cationic exchange-resins after cementation). Preliminary results show the strong effect of chloride on the corrosion susceptibility of the steel. Monitoring will continue for

  11. 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

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

    2017-07-31

    The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity

  13. Regionalization of municipal solid waste management in Japan: balancing the proximity principle with economic efficiency.

    Science.gov (United States)

    Okuda, Itaru; Thomson, Vivian E

    2007-07-01

    The proximity principle - disposing of waste close to its origin - has been a central value in municipal solid waste (MSW) management in Japan for the last 30 years and its widespread adoption has helped resolve numerous "Not in My Backyard" issues related to MSW management. However, MSW management costs have soared, in large part because of aggressive recycling efforts and because most MSW is incinerated in a country that has scarce landfill capacity. In addition, smaller, less sophisticated incinerators have been closed because of high dioxin emissions. Rising costs combined with the closure of smaller incinerators have shifted MSW management policy toward regionalization, which is the sharing of waste management facilities across municipalities. Despite the increased use of regionalized MSW facilities, the proximity principle remains the central value in Japanese MSW management. Municipal solid waste management has become increasingly regionalized in the United States, too, but different driving forces are at work in these two countries. The transition to regionalized MSW management in Japan results from strong governmental control at all levels, with the central government providing funds and policy direction and prefectures and municipalities being the primary implementing authorities. By contrast, market forces are a much stronger force with US MSW management, where local governments - with state government oversight - have primary responsibility for MSW management. We describe recent changes in Japan's MSW programs. We examine the connections between MSW facility regionalization, on the one hand, and, on the other hand, the proximity principle, coordination among local governments, central government control, and financing mechanisms.

  14. The Radioactive Waste Management Advisory Committee's. Advice on issues which need to be addressed in the Guidance to be given to the Environment Agencies on the Principles for determining Radioactive Waste Discharge Authorisations - the 'Principles Document'

    International Nuclear Information System (INIS)

    1998-07-01

    In January 1998, the Minister for the Environment, Mr Michael Meacher, informed the Radioactive Waste Management Advisory Committee (RWMAC) that, during the coming year, he would welcome the Committee's advice on proposals for guidance from the Department of the Environment, Transport and the Regions (DETR) to the Environment Agencies on assessment principles for determining radioactive waste discharge authorisations. This will hereafter be referred to as the Principles Document. The RWMAC has provided advice on the process of regulating radioactive waste discharges for many years. A summary of some of this activity is given in Annex 1. As a result, it has been pressing for this Principles Document guidance to be made available since its Twelfth Report in 1991. In response to the Minister's request, the RWMAC offered to assemble and submit early advice on what it believes the guidance needs to cover: this document fulfils that offer. The fundamental purpose of the advice is to help promote clarity of the regulatory regime for the benefit of the regulators themselves who must apply it, the industry to whom it is applied and, most importantly, the public whose safety it is designed to protect. Clarification of a number of aspects of the process is also likely to provide opportunity for efficiency gains. At a subsequent stage, the RWMAC will be happy to provide comment on any draft principles documentation prepared by the DETR. The RWMAC acknowledges that some of the issues it raises in this advice could be taken by others to be either outside the scope of the Principles Document or, by implying a need for more fundamental consideration of the discharge authorisation process, could potentially preclude its early publication. In the first instance, reference to an alternative source of relevant advice might suffice, providing this advice is itself easily accessible and understandable. In the second, the issue itself might be one to be fed into the Government's planned

  15. 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)

  16. Prediction of waste glass melt rates

    International Nuclear Information System (INIS)

    Lee, L.

    1987-01-01

    Under contract to the Department of Energy, the Du Pont Company has begun construction of a Defense Waste Processing Facility to immobilize radioactive wastes now stored as liquids at the Department of Energy's Savannah River Plant. The immobilization process solidifies waste sludge by vitrification into a leach-resistant borosilicate glass. Development of this process has been the responsibility of the Savannah River Laboratory. As part of the development, a simple model was developed to predict the melt rates for the waste glass melter. This model is based on an energy balance for the cold cap and gives very good agreement with melt rate data obtained from experimental campaigns in smaller scale waste glass melters

  17. Can the same principles be used for the management of radioactive and non-radioactive waste?

    International Nuclear Information System (INIS)

    Bengtsson, Gunnar.

    1989-01-01

    Non-radioactive waste has a much more complex composition than radioactive waste and appears in much larger quantities. The two types of waste have, however, some properties in common when it comes to their longterm impact on health and the environment. The occurrence in both of substances that may exist for generations and may cause cancer provides one example. Both types of waste also always occur together. It is therefore proposed that the same basic principles could be applied for the management of radioactive and non-radioactive waste. By doing so one may increase the efficiency of policy development, research and practical management. This is particurlarly importand for the very costly restoration of old disposal sites which have earlier been poorly managed. (author)

  18. 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

  19. Application of the principles of radioactive waste management to a historical problem

    International Nuclear Information System (INIS)

    Shields, B.; Newbery, S.M.; Finch, A.I.

    1999-01-01

    Faced with a historical radioactive waste situation, the operator has to make several decisions, including: 1. the likely fate of the material, both in the immediate future and long term; 2. the optimum containment for the material so that requirements at all future stages in the material's life can be satisfied; 3. the radiation protection requirements to be met during stages when handling of the waste were required. Americium-24 1, in the form of foils from lightning conductors, had been stored for some 10 years in a vault in Tasmania and required repackaging. This paper discusses: the historical situation, resulting in the need for repackaging; the options available to the operators; the criteria against which options were assessed (including the IAEA principles of radioactive waste management); and the final, practical solution. Copyright (1999) Australasian Radiation Protection Society Inc

  20. 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)

  1. 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.

  2. 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

  3. 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

  4. 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

  5. Nuclear waste management. Quarterly progress report, April-June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Platt, A.M.; Powell, J.A. (comps.)

    1980-09-01

    The status of the following programs is reported: high-level waste immobilization; alternative waste forms; Nuclear Waste Materials Characterization Center; TRU waste immobilization; TRU waste decontamination; krypton solidification; thermal outgassing; iodine-129 fixation; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; mobility of organic complexes of fission products in soils; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology; systems study on engineered barriers; criteria for defining waste isolation; spent fuel and fuel pool component integrity program; analysis of spent fuel policy implementation; asphalt emulsion sealing of uranium tailings; application of long-term chemical biobarriers for uranium tailings; and development of backfill material.

  6. Nuclear waste forms for actinides

    Science.gov (United States)

    Ewing, Rodney C.

    1999-01-01

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

  7. Removing Phosphate from Hanford High-Phosphate Tank Wastes: FY 2010 Results

    Energy Technology Data Exchange (ETDEWEB)

    Lumetta, Gregg J.; Braley, Jenifer C.; Edwards, Matthew K.; Qafoku, Odeta; Felmy, Andrew R.; Carter, Jennifer C.; MacFarlan, Paul J.

    2010-09-22

    The U.S. Department of Energy (DOE) is responsible for environmental remediation at the Hanford Site in Washington State, a former nuclear weapons production site. Retrieving, processing, immobilizing, and disposing of the 2.2 × 105 m3 of radioactive wastes stored in the Hanford underground storage tanks dominates the overall environmental remediation effort at Hanford. The cornerstone of the tank waste remediation effort is the Hanford Tank Waste Treatment and Immobilization Plant (WTP). As currently designed, the capability of the WTP to treat and immobilize the Hanford tank wastes in the expected lifetime of the plant is questionable. For this reason, DOE has been pursuing supplemental treatment options for selected wastes. If implemented, these supplemental treatments will route certain waste components to processing and disposition pathways outside of WTP and thus will accelerate the overall Hanford tank waste remediation mission.

  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. Conditioning and storage of low level radioactive waste in FR Yugoslavia

    International Nuclear Information System (INIS)

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

    2000-01-01

    FR Yugoslavia is a country without any nuclear power plant on its territory. In the last forty years in the country, as a result of the two research reactors operation and also from radionuclides applications in medicine, industry and agriculture, radioactive waste materials of different levels of specific activity are generated. As a temporary solution, these radioactive waste materials are stored in the two interim storage facility. Since one of the storage facilities is completely full with radioactive wastes, packed in metal drums and plastic barrels, and the second one has an effective space for the next few years, attempts are made in the 'Vinca' Institute of Nuclear Sciences in developing the the immobilization process, for low and intermediate level radioactive wastes and their safe disposal. As an immobilization process, cementation process is investigated. Developed immobilization process has, as a final goal, production of solidified waste-matrix mixture form, that is easy for handling and satisfies requirements for interim storage and final disposal. Radioactive wastes immobilized in inactive matrices are to be placed into concrete containers for further manipulation and disposal

  10. 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

  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. 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)

  13. Consolidated waste forms: glass marbles and ceramic pellets

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  14. 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...

  15. 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

  16. 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

  17. Regulatory decision making in the presence of uncertainty in the context of the disposal of long lived radioactive wastes. Third report of the Working group on principles and criteria for radioactive waste disposal

    International Nuclear Information System (INIS)

    1997-10-01

    Plans for disposing of radioactive wastes have raised a number of unique and mostly philosophical problems, mainly due to the very long time-scales which have to be considered. While there is general agreement on disposal concepts and on many aspects of a safety philosophy, consensus on a number of issues remains to be achieved. The IAEA established a subgroup under the International Radioactive Waste Management Advisory Committee (INWAC). The subgroup started its work in 1991 as the ''INWAC Subgroup on Principles and Criteria for Radioactive Waste Disposal''. With the reorganization in 1995 of IAEA senior advisory committees in the nuclear safety area, the title of the group was changed to ''Working Group on Principles and Criteria for Radioactive Waste Disposal''. The working group is intended to provide an open forum for: (1) the discussion and resolution of contentious issues, especially those with an international component, in the area of principles and criteria for safe disposal of waste; (2) the review and analysis of new ideas and concepts in the subject area; (3) establishing areas of consensus; (4) the consideration of issues related to safety principles and criteria in the IAEA's Radioactive Waste Safety Standards (RADWASS) programme; (5) the exchange of information on national safety criteria and policies for radioactive waste disposal. This is the third report of the working group and it deals with the subject of regulatory decision making under conditions of uncertainty which is a matter of concern with respect to disposal of radioactive wastes underground. 14 refs

  18. An Opportunity to Immobilize 1.6 MT or More of Weapons-Grade Plutonium at the Mayak and Krasnoyarsk-26 Sites

    International Nuclear Information System (INIS)

    Jardine, L J; Borisov, G B; Rovny, S I; Kudinov, K G; Shvedov, A A

    2001-01-01

    The Mayak Production Association (PA Mayak), an industrial site in Russia, will be assigned multiple new plutonium disposition missions in order to implement the ''Agreement Between The Government Of The United States Of America And The Government Of Russian Federation Concerning The Management And Disposition Of Plutonium Designated As No Longer Required For Defense Purposes And Related Cooperation'' signed September 1, 2000, by Gore and Kasyanov, In addition, the mission of industrial-scale mixed-oxide (MOX) fabrication will be assigned to either the Mining Chemical Combine (MCC) industrial site at Krasnoyarsk-26 (K-26) or PA Mayak. Over the next decades, these new missions will generate radioactive wastes containing weapons-grade plutonium. The existing Mayak and K-26 onsite facilities and infrastructures cannot currently treat and immobilize these Pu-containing wastes for storage and disposal. However, the wastes generated under the Agreement must be properly immobilized, treated, and managed. New waste treatment and immobilization missions at Mayak may include operating facilities for plutonium metal-to-oxide conversion processes, industrial-scale MOX fuel fabrication, BN-600 PAKET hybrid core MOX fuel fabrication, and a plutonium conversion demonstration process. The MCC K-26 site, if assigned the industrial-scale MOX fuel fabrication mission, would also need to add facilities to treat and immobilize the Pu-containing wastes. This paper explores the approach and cost of treatment and immobilization facilities at both Mayak and K-26. The current work to date at Mayak and MCC K-26 indicates that the direct immobilization of 1.6 MT of weapons-grade plutonium is a viable and cost-effective alternative

  19. 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)

  20. 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.

  1. 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)

  2. Tank waste remediation system: An update

    International Nuclear Information System (INIS)

    Alumkal, W.T.; Babad, H.; Dunford, G.L.; Honeyman, J.O.; Wodrich, D.D.

    1995-02-01

    The US Department of Energy's Hanford Site, located in southeastern Washington State, contains the largest amount and the most diverse collection of highly radioactive waste in the US. High-level radioactive waste has been stored at the Hanford Site in large, underground tanks since 1944. Approximately 217,000 M 3 (57 Mgal) of caustic liquids, slurries, saltcakes, and sludges have accumulated in 177 tanks. In addition, significant amounts of 90 Sr and 137 Cs were removed from the tank waste, converted to salts, doubly encapsulated in metal containers, and stored in water basins. The Tank Waste Remediation System Program was established by the US Department of Energy in 1991 to safely manage and immobilize these wastes in anticipation of permanent disposal of the high-level waste fraction in a geologic repository. Since 1991, significant progress has been made in resolving waste tank safety issues, upgrading Tank Farm facilities and operations, and developing a new strategy for retrieving, treating, and immobilizing the waste for disposal

  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. 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

  5. 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)

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

    International Nuclear Information System (INIS)

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

    1981-09-01

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

  7. Microwave waste processing technology overview

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, R.D.

    1993-02-01

    Applications using microwave energy in the chemical processing industry have increased within the last ten years. Recently, interest in waste treatment applications process development, especially solidification, has grown. Microwave waste processing offers many advantages over conventional waste treatment technologies. These advantages include a high density, leach resistant, robust waste form, volume and toxicity reduction, favorable economics, in-container treatment, good public acceptance, isolated equipment, and instantaneous energy control. The results from the {open_quotes}cold{close_quotes} demonstration scale testing at the Rocky Flats nuclear weapons facility are described. Preliminary results for a transuranic (TRU) precipitation sludge indicate that volume reductions of over 80% are achievable over the current immobilization process. An economic evaluation performed demonstrated cost savings of $11.68 per pound compared to the immobilization process currently in use on wet sludge.

  8. Microwave waste processing technology overview

    International Nuclear Information System (INIS)

    Petersen, R.D.

    1993-02-01

    Applications using microwave energy in the chemical processing industry have increased within the last ten years. Recently, interest in waste treatment applications process development, especially solidification, has grown. Microwave waste processing offers many advantages over conventional waste treatment technologies. These advantages include a high density, leach resistant, robust waste form, volume and toxicity reduction, favorable economics, in-container treatment, good public acceptance, isolated equipment, and instantaneous energy control. The results from the open-quotes coldclose quotes demonstration scale testing at the Rocky Flats nuclear weapons facility are described. Preliminary results for a transuranic (TRU) precipitation sludge indicate that volume reductions of over 80% are achievable over the current immobilization process. An economic evaluation performed demonstrated cost savings of $11.68 per pound compared to the immobilization process currently in use on wet sludge

  9. 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.

  10. 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

  11. 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.

  12. 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

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

    Science.gov (United States)

    Duffó, Gustavo S.; Farina, Silvia B.; Schulz, Fátima M.; Marotta, Francesca

    2010-10-01

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

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

    International Nuclear Information System (INIS)

    Duffo, Gustavo S.; Farina, Silvia B.; Schulz, Fatima M.; Marotta, Francesca

    2010-01-01

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

  15. 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.

  16. 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.

  17. Management of radioactive wastes from nuclear fuels and power plants in Canada

    International Nuclear Information System (INIS)

    Tomlinson, M.; Mayman, S.A.; Tammemagi, H.Y.; Gale, J.; Sanford, B.; Dyne, P.J.

    1977-01-01

    The nature of Canadian nuclear fuel and nuclear generating plant radioactive wastes are summarized. Full exploitation of fission energy resources entails recovery of all fissile and fertile material from spent fuel and separating the fission products as wastes for disposal. A plan for final disposal of all the radioactive wastes is a key component of the waste management scheme. Principles of a scheme for safe, responsible disposal of long-lived radioactive wastes deep underground in isolation from man and the biosphere are outlined. The status of the development and construction program is indicated. We plan to select a site in either a hard rock formation or in a suitable salt bed by 1981 so that a repository can be constructed to begin a demonstration phase in 1986. The repository is to be capable of eventual expansion to accomodate all Canadian nuclear wastes to at least 2050 when in full-scale operation. Extensive geotechnical studies have been initiated in order to select a site, and design and test the repository. We have demonstrated incorporation of fission products in solids that in the short term (17 years) dissolve more slowly than plutonium decays. Investigations of long-term stability are in hand. The principle of retardation of migration of fission products, so that they decay before surfacing, has been tested. Additional capacity for storage of used fuel prior to reprocessing and disposal is required by 1986 and a preliminary design has been prepared for a pool facility to be located at a central fuel recycling and disposal complex. A demonstration of dry storage of fuel in concrete containers is in progress. The quantities of CANDU generating-station wastes and the principles and methods for managing them are summarized. Methods for volume reduction and immobilization by solidification are well advanced. A radioactive-waste operations site is being developed with several different types of surface storage, each with multiple barriers against

  18. Management of radioactive wastes from nuclear fuels and power plants in Canada

    International Nuclear Information System (INIS)

    Tomlinson, M.; Mayman, S.A.; Tammemagi, H.Y.; Gale, J.; Sanford, B.

    1977-01-01

    The nature of Canadian nuclear fuel and nuclear generating plant radioactive wastes is summarized. Full exploitation of fission energy resources entails recovery of all fissile and fertile material from spent fuel and separating the fission products as wastes for disposal. A plan for final disposal of all the radioactive wastes is a key component of the waste management scheme. Principles of a scheme for safe, responsible disposal of long-lived radioactive wastes deep underground, in isolation from man and the biosphere, are outlined. The status of the development and construction programme is indicated. It is planned to select a site in either a hard rock formation or in a suitable salt bed by 1981 so that a repository can be constructed to begin a demonstration phase in 1986. The repository is to be capable of eventual expansion to accomodate all Canadian nuclear wastes to at least 2050 when in full-scale operation. Extensive geotechnical studies have been initiated in order to select a site, and design and test the repository. The incorporation of fission products in solids that in the short term (17 years) dissolve more slowly than plutonium decays has been demonstrated. Investigations of long-term stability are in hand. The principle of retardation of migration of fission products, so that they decay before surfacing, has been tested. Additional capacity for storage of used fuel prior to reprocessing and disposal is required by 1986 and a preliminary design has been prepared for a pool facility to be located at a central fuel recycling and disposal complex. A demonstration of dry storage of fuel in concrete containers is in progress. The quantities of CANDU generating-station wastes and the principles and methods for managing them are summarized. Methods for volume reduction and immobilization by solidification are well advanced. A radioactive-waste operations site is being developed with several different types of surface storage, each with multiple barriers

  19. 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

  20. 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 ...

  1. 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

  2. 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

  3. Physical immobilization of biofunctional substance by the use of radiation polymerization

    International Nuclear Information System (INIS)

    Kobayashi, M.; Kaetsu, I.

    1982-01-01

    Radiation-induced polymerization at low temperatures of glass-forming monomers in a supercooled state can be applied, for example, in casting of organic glass and the immobilization of biofunctional substance. The immobilization of various biofunctional materials such as enzymes, microbial cells, tissue cells etc. will be a promising application in the near future in biotechnology and bioengineering. The authors studied the immobilization technique which can be applied to general biocomponents, by using low-temperature radiation polymerization in a supercooled phase. According to this method, biocomponents are composed mainly on the surface of the carrier polymer, and therefore the product has the bioactivity at the surface of the composite. This method can be called the adhesion method. Biocomponents can be composed simply by mixing with monomer, shaping into a desirable form, then cooling to low temperature and irradiating into a product. On the cooling of the monomer-buffer (including biocomponent) mixture, water in the buffer changes to ice and then the biocomponents in the buffer are isolated from the ice, and concentrated on a surface of supercooled monomer phase. These biocomponents are fixed immediately on the polymer surface by irradiation. Anti-cancers immobilized by low-temperature radiation-induced polymerization have been applied to local chemotherapy by implantation, and the result of such a slow release system has been proved to be successful by animal experiments. The application of the radiation immobilized antibody to immunoassay has also been proved successful. The authors started research on the utilization of radiation techniques for the conversion of cellulosic wastes such as chaff, rice straw, sawdust, bagasse and wastepaper. It includes pretreatment by irradiation of cellulose wastes and saccharification and fermentation by using radiation immobilized enzymes and yeasts. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

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

    International Nuclear Information System (INIS)

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

    1996-12-01

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

  6. Comparison of SRP high-level waste disposal costs for borosilicate glass and crystalline ceramic waste forms

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-04-01

    An evaluation of costs for the immobilization and repository disposal of SRP high-level wastes indicates that the borosilicate glass waste form is less costly than the crystalline ceramic waste form. The wastes were assumed immobilized as glass with 28% waste loading in 10,300 reference 24-in.-diameter canisters or as crystalline ceramic with 65% waste loading in either 3400 24-in.-diameter canisters or 5900 18-in.-diameter canisters. After an interim period of onsite storage, the canisters would be transported to the federal repository for burial. Total costs in undiscounted 1981 dollars of the waste disposal operations, excluding salt processing for which costs are not yet well defined, were about $2500 million for the borosilicate glass form in reference 24-in.-diameter canisters, compared to about $2900 million for the crystalline ceramic form in 24-in.-diameter canisters and about $3100 million for the crystalline ceramic form in 18-in.-diameter canisters. No large differences in salt processing costs for the borosilicate glass and crystalline ceramic forms are expected. Discounting to present values, because of a projected 2-year delay in startup of the DWPF for the crystalline ceramic form, preserved the overall cost advantage of the borosilicate glass form. The waste immobilization operations for the glass form were much less costly than for the crystalline ceramic form. The waste disposal operations, in contrast, were less costly for the crystalline ceramic form, due to fewer canisters requiring disposal; however, this advantage was not sufficient to offset the higher development and processing costs of the crystalline ceramic form. Changes in proposed Nuclear Regulatory Commission regulations to permit lower cost repository packages for defense high-level wastes would decrease the waste disposal costs of the more numerous borosilicate glass forms relative to the crystalline ceramic forms

  7. 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

  8. The immobilization of filamentous fungi by cellular adherence and surface film growth on radiopolymerized hydrogels

    International Nuclear Information System (INIS)

    Petre, M.; Zarnea, G.; Adrian, P.; Gheorghiu, E.; Gheordunescu, V.

    1998-01-01

    The continuous biodegradation processes of cellulose wastes, from wine-producing industry and canned food processing were performed using immobilized fungal cells on complex polyhydrogels such as: collagen-polyacrylamide (CPAA), elastin-polyacrylamide (EPAA), gelatin-polyacrylamide (GPAA) and poly-hydroxy-ethyl-methyl-methacrylate (PHEMA) which were prepared by experimental radiopolymerization methods using a 60 Co radioactive source. The experiments were performed using cellulolytic micro- and macrofungal species and the immobilization methods were applied in connection with morphogenetic characteristics of each one of these species that were used as inoculum. A flow-cascade bioreactor was achieved as a research laboratory tool in order to control the cellulose waste biodegradation in continuous processes using immobilized fungal cells on complex polyhydrogels. In this way the continuous control was possible for fungal metabolic activity during the cellulose biodegradation processes by microscopical hyphal growth measurements inside the polyhydrogels and, also, by mycelial protein content analysis. The tested fungal species were Pleurotus ostreatus and Pleurotus florida from macromycetes and Trichoderma viride from micromycetes. There were noticed significant differences in biodegradation rate of cellulose wastes between 'in vitro' immobilized fungal cultures and free fungal cells cultures used as control samples to be compared with cellulose biodegradation processes in natural conditions of polluted ecosystems. The fungal growth displayed by soluble mycelial protein were compared with cellulose biodegradation rates which were expressed by cellulose weight loss in the culture medium. (authors)

  9. 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

  10. 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.)

  11. 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

  12. Studies on the immobilization of simulated HLW in NaTi2(PO4)3 (NTP) matrix

    International Nuclear Information System (INIS)

    Raja Madhavan, R.; Govindan Kutty, K.V.; Gandhi, A.S.

    2015-01-01

    Immobilization of high level nuclear waste (HLW) is a big challenge faced by the nuclear industry today. The HLW has to be contained and isolated from the biosphere for geological timescales. NZP family of compounds is very versatile monophasic hosts for HLW immobilization. Their crystal structure can accommodate nearly all the cations known to be present in HLW due to its open structure with voids of different size. In the present study a systematic investigation on NaTi 2 (PO 4 ) 3 belonging to the NZP family; as a potential host for HLW immobilization was carried out. A simulated HLW expected from Fast Breeder Test Reactor, India (FBTR) (150Gwd/T burnup, 1 year cooling) was used. Simulated NTP waste forms with 5, 10, 15 wt. % waste loading were prepared by employing a wet chemical method and characterized. Single phase simulated NTP waste forms with up to 5 wt.% waste loading could be prepared for samples sintered in air and above 5 wt.% waste loading, monazite phase is observed as a minor secondary phase. It was found that when sintering is done in Ar/10%H 2 , NTP matrix accepts up to 10 wt.% waste loading without formation of any second phase. From the SEM studies, it was observed that samples sintered in air as well as Ar/10%H 2 palladium segregated as a metal phase and uniformly distributed throughout the waste matrix. The elemental mapping revealed retention of some of the fission products like Ru, Mo, Cs that are volatile during sintering above 1173 K and are homogenously distributed in the matrix. (author)

  13. Optimization of concrete composition in radioactive waste management

    International Nuclear Information System (INIS)

    Plecas, I.; Peric, A.

    1995-01-01

    Low and intermediate level waste represents 95% of the total wastes that is conditioned into special concrete containers. Since these containers are to protect radioactive waste safely for about 300 years, the selection and precise control of physical and mechanical characteristics of materials is very important. After volume reduction and valuable components recovery, waste materials have to be conditioned for transport, storage and disposal. Conditioning is the waste management step in which radioactive wastes are immobilized and packed. The immobilization processes involve conversation of the wastes to solid forms that reduce the potential for migration or dispersion of radionuclides from the wastes by natural processes during storage, transport and disposal. The immobilization processes involve the use of various matrices of nonradioactive materials, such as concrete, to fix the wastes as monoliths, usually directly in the waste containers used for subsequent handling. In this paper an optimization of concrete container composition, used for storing radioactive waste from nuclear power plants, is presented. Optimization was performed on the composition of the concrete that is used in the container production. In experiments, the authors tried to obtain the best mechanical characteristics of the concrete, varying the weight percentage of the granulate due to its diameter, water-to-cement ratios and type of the cements that were used in preparing the concrete container formulation. Concrete containers, that were optimized in the manner described in this paper, will be in used for the radioactive waste materials final disposal, using the concept of the engineer trench system facilities

  14. Vitrification of low level and mixed (radioactive and hazardous) wastes: Lessons learned from high level waste vitrification

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1994-01-01

    Borosilicate glasses will be used in the USA and in Europe immobilize radioactive high level liquid wastes (HLLW) for ultimate geologic disposal. Simultaneously, tehnologies are being developed by the US Department of Energy's (DOE) Nuclear Facility sites to immobilize low-level and mixed (radioactive and hazardous) wastes (LLMW) in durable glass formulations for permanent disposal or long-term storage. Vitrification of LLMW achieves large volume reductions (86--97 %) which minimize the associated long-term storage costs. Vitrification of LLMW also ensures that mixed wastes are stabilized to the highest level reasonably possible, e.g. equivalent to HLLW, in order to meet both current and future regulatory waste disposal specifications The tehnologies being developed for vitrification of LLMW rely heavily on the technologies developed for HLLW and the lessons learned about process and product control

  15. 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 ...

  16. 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)

  17. Solid waste management. Principles and practice

    Energy Technology Data Exchange (ETDEWEB)

    Chandrappa, Ramesha [Karnataka State Pollution Control Board, Biomedical Waste, Bangalore (India); Bhusan Das, Diganta [Loughborough Univ. of Technology (United Kingdom). Dept. of Chemical Engineering

    2012-11-01

    Solid waste was already a problem long before water and air pollution issues attracted public attention. Historically the problem associated with solid waste can be dated back to prehistoric days. Due to the invention of new products, technologies and services the quantity and quality of the waste have changed over the years. Waste characteristics not only depend on income, culture and geography but also on a society's economy and, situations like disasters that affect that economy. There was tremendous industrial activity in Europe during the industrial revolution. The twentieth century is recognized as the American Century and the twenty-first century is recognized as the Asian Century in which everyone wants to earn 'as much as possible'. After Asia the currently developing Africa could next take the center stage. With transitions in their economies many countries have also witnessed an explosion of waste quantities. Solid waste problems and approaches to tackling them vary from country to country. For example, while efforts are made to collect and dispose hospital waste through separate mechanisms in India it is burnt together with municipal solid waste in Sweden. While trans-boundary movement of waste has been addressed in numerous international agreements, it still reaches developing countries in many forms. While thousands of people depend on waste for their lively hood throughout the world, many others face problems due to poor waste management. In this context solid waste has not remained an issue to be tackled by the local urban bodies alone. It has become a subject of importance for engineers as well as doctors, psychologist, economists, and climate scientists and any others. There are huge changes in waste management in different parts of the world at different times in history. To address these issues, an effort has been made by the authors to combine their experience and bring together a new text book on the theory and practice of the

  18. Nuclear waste disposal: perspective of a geochemist

    International Nuclear Information System (INIS)

    Sengupta, Pranesh; Dey, G.K.

    2011-01-01

    Satisfying the growing requirement in an environment friendly way is one of the most important tasks we need to accomplish these days. Considering the restricted non-renewable energy resources and limited technological progresses achieved in the renewable energy sectors in India, nuclear energy appears to be one of the most lucrative solutions towards the forthcoming energy crisis. Successful implementation of nuclear energy program however requires careful execution of high level nuclear waste management activities. One very important aspect of this process is to identify and develop suitable inert matrix(ces) for conditioning of nuclear waste(s) using natural analogue studies. And this establishes the very vital linkage between geochemical studies and nuclear waste immobilization. One good example of such an interdisciplinary approach can be seen in the methodologies adopted for immobilization of sulfate bearing high level nuclear wastes (SO 4 -HLW). It has been reported on several occasions that sulfur-rich melt get separated from silicate melt within magma chamber. Similar process has also been witnessed within vitrification furnaces whenever an attempt has been made to condition SO 4 -HLW within borosilicate glass matrices. Since such liquid-liquid phase separation leads to multiple difficulties in connection to radionuclide immobilization and plant scale vitrification processes, solutions were sought from natural analogue studies. Such as integrated approach ultimately resulted in establishing two different methodologies e.g. (i) modifying the borosilicate network through introduction of Ba 2+ cation; a process being followed in India and (ii) using phosphatic melt as a host instead of borosilicate melt; a process being followed in Russia. Detail of these two routes and the geochemical linkage in nuclear waste immobilization will be discussed.(author)

  19. Low-temperature setting phosphate ceramics for stabilization of DOE problem low level mixed-waste: I. Material and waste form development

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.; Knox, L.; Mayberry, J.

    1994-03-01

    Chemically bonded phosphate ceramics are proposed as candidates for solidification and stabilization of some of the open-quotes problemclose quotes DOE low-level mixed wastes at low-temperatures. Development of these materials is crucial for stabilization of waste streams which have volatile species and any use of high-temperature technology leads to generation of off-gas secondary waste streams. Several phosphates of Mg, Al, and Zr have been investigated as candidate materials. Monoliths of these phosphates were synthesized using chemical routes at room or slightly elevated temperatures. Detailed physical and chemical characterizations have been conducted on some of these phosphates to establish their durability. Magnesium ammonium phosphate has shown to possess excellent mechanical and as well chemical properties. These phosphates were also used to stabilize a surrogate ash waste with a loading ranging from 25-35 wt.%. Characterization of the final waste forms show that waste immobilization is due to both chemical stabilization and physical encapsulation of the surrogate waste which is desirable for waste immobilization

  20. Immobilization of {sup 99}Tc (Re) using Iron-Phosphate Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Jong; Xu, Kai; Um, Woo Yong; Hrma, Pavel [Pohang Univ. of Science and Technology, Pohang (Korea, Republic of)

    2012-05-15

    Technetium-99 ({sup 99}Tc) is a fission product artificially generated during the irradiation of {sup 235}U for commercial power production or {sup 239}Pu for nuclear weapons. Under oxidizing conditions, the dominant species of Tc, the pertechnetate anion (TcO{sub 4} {sup -}), is highly soluble in ground water and thus easily transports through the geologic systems. In addition, because of its high fission yield ({approx}6 %) and long half-life (2.1x10{sup 5} yr), immobilization of {sup 99}Tc has been investigated for decades. Several waste forms such as metallic alloys, sintered titanate ceramics and chemically bonded phosphate ceramics have been proposed to encapsulate {sup 99}Tc. They have not yet been realized in the industrial-scale, mostly either due to the high volatilization of {sup 99}Tc during high temperature process (>1300 .deg. C), or the low {sup 99}Tc loading. Iron-phosphate (FeP) glasses have been developed as alternative waste forms because of their chemical durability equivalent to borosilicate glasses. Additionally, vitrification of radioactive waste by FeP glasses can be done at a relatively low temperature ({approx}1000 .deg. C) and the low-temperature process can reduce the volatilization of {sup 99}Tc significantly. Thus, this work reports the immobilization of {sup 99}Tc by FeP glasses using rhenium (Re) as a surrogate. We also examine the chemical durability of Re-containing FeP glasses using product consistency test (PCT). Experimental results reveal that FeP glass can become a promising candidate for immobilizing {sup 99}Tc

  1. Small-scale integrated demonstration of high-level radioactive waste processing and vitrification using actual SRP waste

    International Nuclear Information System (INIS)

    Ferguson, R.B.; Woolsey, G.B.; Galloway, R.M.; Baumgarten, P.M.; Eibling, R.E.

    1980-01-01

    Experiments have been made to demonstrate the feasibility of immobilizing SRP high-level waste in borosilicate glass. Results to date are encouraging. Equipment performance and processing characteristics for solidifying small batches of actual SRP waste have agreed well with previous experience with small- and large-scale tests synthetic waste, and with theoretical predictions

  2. Microwave process employed to study the immobilization feasibility of spent ion exchange resins in polymeric matrices

    Energy Technology Data Exchange (ETDEWEB)

    Caratin, Reinaldo L.; Araujo, Sumair G. de; Landini, Liliane; Neves, Sabrina C.; Lugao, Ademar B. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], E-mail: rcaratin@ipen.br, E-mail: sgaraujo@ipen.br, E-mail: llandini@ipen.br, E-mail: scneves@ipen.br, E-mail: ablugao@ipen.br

    2007-07-01

    Nuclear activities generate radioactive wastes in several physical states, radioactive levels and kinds of radioactive emission. Hence, a lot of techniques have been developed and optimized to do the immobilization of these materials, according to local and international regulations to protect human being and environment. Another great concern is the indiscriminate disposal of used polymeric materials (such as plastic and rubber) or production leftovers in landfills, which remain for many years before they are naturally decomposed. In this work, it was studied the possibility of carrying out the immobilization of spent ion exchange resins (contaminated with ionising radiation), by using polymeric matrices of bitumen and rubber (as solidification materials for the storage of low level radioactive waste). The samples were mixed at different percentages and were heated in a microwave device (2,450 MHz) at IPEN/CNEN-SP, varying the irradiation time and power. The objective of the immobilization is converting the wastes into forms that are leach resistant and physically and chemically stable for disposal. Characterizations of these materials have been performed according to ABNT-NBR standards. The results indicated the previous idea of the necessary minimum temperature to keep the matrix for future embedding of radioactive waste, in solid state. (author)

  3. 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

  4. 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.

  5. Waste Management Strategy in The Netherlands. Part 2. Strategy Principles and Influencing Issues

    International Nuclear Information System (INIS)

    Haverkate, B.R.W.

    2002-01-01

    This report reflects the Dutch input prepared in the framework of work package 2 of the EU thematic network COMPAS, which deals with the identification of alternative waste management strategies and issues influencing strategy selection in EU member states and their applicant countries. All elements that could have an effect in identifying alternative policies to manage (long-lived) radioactive wastes are addressed in this report. After a short introduction, in chapter 1, about some general issues influencing decision-making such as public acceptance, involvement, perception and (European) legislation, the considered disposal methods and disposal requirements are given in chapter 2. Chapter 3 of this report deals with the background topics of the current waste management strategy in The Netherlands. A detailed overview of (basic) strategy principles and their influencing issues is the subject of chapter 4. Issues considered include: safety and environmental impact; technical limitations; nuclear materials safeguards; monitoring and retrievability; ethical issues; public acceptance; (timing of) strategy development and implementation; and economical considerations. Relevant additional issues that could have an effect in identifying alternative waste management strategy are provided in appendices, including signed treaties (appendix B) and nuclear statutory regulations (appendix C)

  6. Use of diatomaceous to liquid organic wastes adsorption

    International Nuclear Information System (INIS)

    Sanhueza M, Azucena; Padilla S, Ulises

    1999-01-01

    Background: One of the radioactive wastes that the Radioactive Wastes Management Unit must process are organic liquids from external generators and from sections of the Chilean Nuclear Energy Commission (CCHEN). The wastes from external generators contain H 3 and C 14; while the wastes from the CCHEN are contaminated with uranium. The total volume of liquid organic wastes that must be treated is 5 m3. The options recommended for processing these wastes are incineration or the adsorption of the organic liquid by some adsorbing medium and its subsequent immobilization in cement molds. Due to the cost of incineration, the adsorption method was chosen for study. Objective: To find the optimum amount of adsorbent to be saturated with radioactive organic liquid from liquid scintillation and to study immobilization in cement molds. Methodology: Adsorption granulated (1568 Merck) and diatom earth were tested as adsorbent mediums. The adsorbents were mixed in different ratios of volume with the organic liquid. Then the waste was mixed with different water/cement ratios to define the best immobilization conditions. Conclusions: The tests carried out with 2 adsorbents recommended in the literature and available in the CCHEN show that as adsorbent waste ratio decreases, the percentage of liquid adsorbed increases, as expected: a greater volume of adsorbent retains a greater quantity of liquid, with an increase in the final volume, depending on the adsorbent used. Of these adsorbents, the diatom earth was better for treating liquid organic wastes. It had 100% adsorption and an increased volume of 0%, which is more than enough from the volumetric point of view of waste management. The ratio 0.8 liquid/adsorbent also showed good characteristics, but more study is needed to decide on the above, since liquid remains to be adsorbed. This work must continue to study the repeatability of results, to obtain physical and radiological characteristics for the immobilized products and to

  7. 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

  8. Separating and stabilizing phosphate from high-level radioactive waste: process development and spectroscopic monitoring.

    Science.gov (United States)

    Lumetta, Gregg J; Braley, Jenifer C; Peterson, James M; Bryan, Samuel A; Levitskaia, Tatiana G

    2012-06-05

    Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

  9. 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

  10. 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.

  11. Radiation Effects in Nuclear Waste Materials

    International Nuclear Information System (INIS)

    Weber, William J.; Wang, Lumin; Hess, Nancy J.; Icenhower, Jonathan P.; Thevuthasan, Suntharampillai

    2003-01-01

    The objective of this project is to develop a fundamental understanding of radiation effects in glasses and ceramics, as well as the influence of solid-state radiation effects on aqueous dissolution kinetics, which may impact the performance of nuclear waste forms and stabilized nuclear materials. This work provides the underpinning science to develop improved glass and ceramic waste forms for the immobilization and disposition of high-level tank waste, excess plutonium, plutonium residues and scrap, other actinides, and other nuclear waste streams. Furthermore, this work is developing develop predictive models for the performance of nuclear waste forms and stabilized nuclear materials. Thus, the research performed under this project has significant implications for the immobilization of High-Level Waste (HLW) and Nuclear Materials, two mission areas within the Office of Environmental Management (EM). With regard to the HLW mission, this research will lead to improved understanding of radiation-induced degradation mechanisms and their effects on dissolution kinetics, as well as development of predictive models for waste form performance. In the Nuclear Materials mission, this research will lead to improvements in the understanding of radiation effects on the chemical and structural properties of materials for the stabilization and long-term storage of plutonium, highly-enriched uranium, and other actinides. The research uses plutonium incorporation, ion-beam irradiation, and electron-beam irradiation to simulate the effects of alpha decay and beta decay on relevant glasses and ceramics. The research under this project has the potential to result in improved glass and ceramic materials for the stabilization and immobilization of high-level tank waste, plutonium residues and scraps, surplus weapons plutonium, highly-enriched uranium, other actinides, and other radioactive materials

  12. Radiation Effects in Nuclear Waste Materials

    International Nuclear Information System (INIS)

    Weber, William J.

    2005-01-01

    The objective of this project is to develop a fundamental understanding of radiation effects in glasses and ceramics, as well as the influence of solid-state radiation effects on aqueous dissolution kinetics, which may impact the performance of nuclear waste forms and stabilized nuclear materials. This work provides the underpinning science to develop improved glass and ceramic waste forms for the immobilization and disposition of high-level tank waste, excess plutonium, plutonium residues and scrap, other actinides, and other nuclear waste streams. Furthermore, this work is developing develop predictive models for the performance of nuclear waste forms and stabilized nuclear materials. Thus, the research performed under this project has significant implications for the immobilization of High-Level Waste (HLW) and Nuclear Materials, two mission areas within the Office of Environmental Management (EM). With regard to the HLW mission, this research will lead to improved understanding of radiation-induced degradation mechanisms and their effects on dissolution kinetics, as well as development of predictive models for waste form performance. In the Nuclear Materials mission, this research will lead to improvements in the understanding of radiation effects on the chemical and structural properties of materials for the stabilization and long-term storage of plutonium, highly-enriched uranium, and other actinides. The research uses plutonium incorporation, ion-beam irradiation, and electron-beam irradiation to simulate the effects of alpha decay and beta decay on relevant glasses and ceramics. The research under this project has the potential to result in improved glass and ceramic materials for the stabilization and immobilization of high-level tank waste, plutonium residues and scraps, surplus weapons plutonium, highly-enriched uranium, other actinides, and other radioactive materials

  13. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Rummery, T.E.; Rosinger, E.L.J.

    1983-05-01

    The Canadian Nuclear Fuel Waste Management Program is now well established. This report outlines the generic research and technological development underway in this program to assess the concept of immobilization and subsequent disposal of nuclear fuel waste deep in a stable plutonic rock in the Canadian Shield. The program participants, funding, schedule and associated external review processes are briefly outlined. The major scientific and engineering components of the program, namely, immobilization studies, geoscience research and environmental and safety assessment, are described in more detail

  14. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Rummery, T.E.; Rosinger, E.L.J.

    1984-12-01

    The Canadian Nuclear Fuel Waste Management Program is 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 basic safety and environmental aspects of the concept of isolating immobilized fuel waste by deep underground disposal in plutonic rock. The major scientific and engineering components of the program, namely immobilization studies, geoscience research, and environmental and safety assessment, are described. Program funding, scheduling and associated external review processes are briefly outlined

  15. 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)

  16. Radioactive demonstration of final mineralized waste forms for Hanford waste treatment plant secondary waste (WTP-SW) by fluidized bed steam reforming (FBSR) using the bench scale reformer platform

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Burket, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Cozzi, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jantzen, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Missimer, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-08-01

    The U.S. Department of Energy’s Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford’s tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as 137Cs, 129I, 99Tc, Cl, F, and SO4 that volatilize at the vitrification temperature of 1150°C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW.

  17. Iron phosphate glass: a promising matrix for the immobilization of Cs and Mo

    International Nuclear Information System (INIS)

    Hemadevi, V.; Joseph, Kitheri

    2015-01-01

    Presently, borosilicate glass (BSG) is the acceptable vitrification matrix for the immobilization of high level waste. The solubility of Mo in BSG is limited in the presence of Cs. As per the literature, solubility of Mo in BSG is about 2.5 wt. % in the presence of Cs. Hence it is difficult to immobilize nuclear waste rich in Cs and Mo in borosilicate glass. It is observed that the composition of Cs and Mo expressed as oxides are 10.4 and 14.7 wt. % respectively in simulated fast reactor waste. Iron phosphate glass containing 20 wt. % simulated fast reactor waste (referred as IP20FRW) was synthesized and characterized. IP20FRW contains ~ 3 wt. % of molybdenum oxide along with 2 wt. % cesium oxide. IPG is a suitable matrix for the immobilization of Cs and Mo separately. Hence it is essential to understand the glass characteristics of IPG containing both Cs and Mo. This paper explores systematic loading of both Cs and Mo such that the final composition corresponds to 10.5 wt. % Cs 2 O-15 wt. % MoO 3 -31.9 wt. % Fe 2 O 3 -42.6 wt. % P 2 O 5 . In addition to synthesis, the present study also includes understanding the change in glass characteristics of IPG containing both Cs and Mo. The possibility of higher percent loading of both Cs and Mo in IPG demonstrates the better glass forming characteristics of IPG. The synthesis and characterization of Cs-Mo loaded glasses will be discussed in this paper. (author)

  18. 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

  19. 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

  20. Radioactive wastes management development in Chile

    International Nuclear Information System (INIS)

    Mir, S.A.; Cruz, P.F.; Rivera, J.D.; Jorquera, O.H.

    1994-01-01

    A Facility for immobilizing and conditioning of radioactive wastes generated in Chile, has recently started in operation. It is a Radioactive Wastes Treatment Plant, RWTP, whose owner is Comision Chilena de Energia Nuclear, CCHEN. A Storgement Building of Conditioned Wastes accomplishes the facility for medium and low level activity wastes. The Project has been carried with participation of chilean professionals at CCHEN and Technical Assistance of International Atomic Energy Agency, IAEA. Processes developed are volume reduction by compaction; immobilization by cementation and conditioning. Equipment has been selected to process radioactive wastes into a 200 liters drum, in which wastes are definitively conditioned, avoiding exposition and contamination risks. The Plant has capacity to treat low and medium activity radioactive wastes produced in Chile due to Reactor Experimental No. 1 operation, and annex Laboratories in Nuclear Research Centers, as also those produced by users of nuclear techniques in Industries, Hospitals, Research Centers and Universities, in the whole country. With the infrastructure developed in Chile, a centralization of Radioactive Wastes Management activities is achieved. A data base system helps to control and register radioactive wastes arising in Chile. Generation of radioactive wastes in Chile, has found solution for the present production and that of near future

  1. Current technics and management strategy for Pu-contaminated wastes at PNC

    International Nuclear Information System (INIS)

    1981-02-01

    Power Reactor and Nuclear Fuel Development Corporation (PNC) was designated as a leading organization for the Pu-contaminated waste technology program in Japan. For this purpose, number of efforts in the research and development are proceeding. That is, Pu-contaminated waste technology including volume reduction system and the immobilization of wastes is being developed. The design of a Pu-contaminated waste treatment facility (PWTF) is being made for the demonstration of the technology developed. Studies are in progress to find the criteria for waste products in disposal. The current procedures and strategy for the management of Pu-contaminated wastes at PNC are described as follows: current and future management; technology development including controlled air incineration, acid digestion, immobilization melting, dismantling, and liquid waste treatment; the Pu-contaminated waste treatment facility. (J.P.N.)

  2. Defense Waste Processing Facility, Savannah River Plant

    International Nuclear Information System (INIS)

    After 10 years of research, development, and testing, the US Department of Energy is building a new facility which will prepare high-level radioactive waste for permanent disposal. The Defense Waste Processing Facility, known as the DWPF, will be the first production-scale facility of its kind in the United States. In the DWPF, high-level waste produced by defense activities at the Savannah River Plant will be processed into a solid form, borosilicate glass, suitable for permanent off-site geologic disposal. With construction beginning in the fall of 1983, the DWPT is scheduled to be operational in 1989. By 2005, the DWPF will have immobilized the backlog of high-level waste which has been accumulating in storage tanks at the Savannah River Plant since 1954. Canisters of the immobilized waste will then be ready for permanent disposal deep under the ground, safely isolated from the environment

  3. 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.

  4. Stepwise approach to decision making for long-term radioactive waste management. Experience, issues and guiding principles

    International Nuclear Information System (INIS)

    2004-01-01

    Radioactive waste exists as a result of both past and current practices. One of the most challenging tasks is the management of long-lived waste that must be isolated from the human environment for many thousands, or even hundreds of thousands, of years. Although significant technical progress has been made in developing management schemes that, according to technical experts, would ensure long-term safety (e.g. engineered geologic disposal), the rate of progress towards implementing such solutions has been slower than expected. The contrast between expected and observed rates may be partly attributable to an earlier technical optimism. More significant, however, are the setbacks, which have arisen mainly from an underestimation of the societal and political dimensions. In long-term radioactive waste management, consideration is increasingly being given to concepts such as stepwise decision making and adaptive staging in which the public, and especially the local public, are to be meaningfully involved in the review and planning of developments. The key feature of these concepts is development by steps or stages that are reversible, within the limits of practicability. This is designed to provide reassurance that decisions can be reversed if experience shows them to have adverse or unwanted effects. A stepwise approach to decision making has thus come to the fore as being of value in advancing long-term radioactive waste management solutions in a societally acceptable manner. Despite its early identification within the radioactive waste management community as an important means for reaching solutions and decisions in which there is broad-based confidence, the bases for and application of stepwise decision making has not been widely reviewed. Guiding principles of any such process are still being formulated, its roots in empirical social science research have not been fully reviewed, nor the difficulties of its implementation analysed. The report reviews current

  5. Purification and solidification of reactor wastes at a Canadian nuclear generating station

    International Nuclear Information System (INIS)

    Buckley, L.P.; Burt, D.A.

    1981-06-01

    Chalk River Nuclear Laboratories are developing methods to condition power reactor wastes and to immobilize their radionuclides. Evaporation alone and combined with bituminization has been an important part of the program. After testing at the laboratories a 0.5 m 2 wiped-film evaporator was sent to the Douglas Point Nuclear Generating Station (220 MWe) to demonstrate its suitability to handle typical reactor liquid wastes. Two specific tasks undertaken with the wiped-film evaporator were successfully completed. The first was purification of contaminated heavy water which had leaked from the moderator circuit. The heavy water is normally recovered, cleaned by filters and ion-exchange resin and then upgraded by electrolysis. Cleaning the heavy water with the wiped-film evaporator produced better quality water for upgrading than had been achieved by any previous method and at much lower operating cost. The second task was to concentrate and immobilize a decontamination waste. The waste was generated from the decontamination of pump bowls used in the primary heat transport circuit. The simultaneous addition of the liquid waste and bitumen emulsion to the wiped-film evaporator produced a solid containing 30 wt% waste solids in a bitumen matrix. The volume reduction achieved was 16:1 based on the volumes of initial liquid waste and the final product generated. The quantity sent to storage was 20 times less than had the waste been immobilized in a cement matrix. The successful demonstration has resulted in a proposal to install a wiped-film evaporator at the station to clean heavy water and immobilize decontamination wastes. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  7. Optimizing Urban Material Flows and Waste Streams in Urban Development through Principles of Zero Waste and Sustainable Consumption

    Directory of Open Access Journals (Sweden)

    Steffen Lehmann

    2011-01-01

    Full Text Available Beyond energy efficiency, there are now urgent challenges around the supply of resources, materials, energy, food and water. After debating energy efficiency for the last decade, the focus has shifted to include further resources and material efficiency. In this context, urban farming has emerged as a valid urban design strategy, where food is produced and consumed locally within city boundaries, turning disused sites and underutilized public space into productive urban landscapes and community gardens. Furthermore, such agricultural activities allow for effective composting of organic waste, returning nutrients to the soil and improving biodiversity in the urban environment. Urban farming and resource recovery will help to feed the 9 billion by 2050 (predicted population growth, UN-Habitat forecast 2009. This paper reports on best practice of urban design principles in regard to materials flow, material recovery, adaptive re-use of entire building elements and components (‘design for disassembly’; prefabrication of modular building components, and other relevant strategies to implement zero waste by avoiding waste creation, reducing wasteful consumption and changing behaviour in the design and construction sectors. The paper touches on two important issues in regard to the rapid depletion of the world’s natural resources: the built environment and the education of architects and designers (both topics of further research. The construction and demolition (C&D sector: Prefabricated multi-story buildings for inner-city living can set new benchmarks for minimizing construction wastage and for modular on-site assembly. Today, the C&D sector is one of the main producers of waste; it does not engage enough with waste minimization, waste avoidance and recycling. Education and research: It’s still unclear how best to introduce a holistic understanding of these challenges and to better teach practical and affordable solutions to architects, urban

  8. A comparison of high-level waste form characteristics

    International Nuclear Information System (INIS)

    Salmon, R.; Notz, K.J.

    1991-01-01

    The US DOE is responsible for the eventual disposal in a repository of spent fuels, high-level waste (HLW) and other radioactive wastes that may require long-term isolation. This includes light-water reactor (LWR) spent fuel and immobilized HLW as the two major sources, plus other forms including non-LWR spent fuels and miscellaneous sources (such as activated metals in the Greater-Than-Class-C category). The Characteristics Data Base, sponsored by DOE's Office of Civilian Radioactive Waste Management (OCRWM), was created to systematically tabulate the technical characteristics of these materials. Data are presented here on the immobilized HLW forms that are expected to be produced between now and 2020

  9. Cyprodinil retention on mixtures of soil and solid wastes from wineries. Effects of waste dose and ageing

    DEFF Research Database (Denmark)

    Rodriguez-Salgado, I.; Paradelo Pérez, Marcos; Pérez-Rodríguez, P.

    2014-01-01

    In spite of its wide-world economic relevance, wine production generates a huge amount of waste that threatens the environment. A batch experiment was designed to assess the effect of the amendment of an agricultural soil with two winery wastes (perlite and bentonite wastes) in the immobilization...... increased. The use of these winery wastes contributes to a more sustainable agriculture preventing fungicide mobilization to groundwater....

  10. Efficient production of succinic acid in immobilized fermentation with crude glycerol from Escherichia coli

    Directory of Open Access Journals (Sweden)

    Nik Nor Aziati, A.A.

    2017-10-01

    Full Text Available The increase in the price of commercial succinic acid has necessitated the need for its synthesis from waste materials such as glycerol. Glycerol residue is a waste product of Oleochemical production which is cheaply available and a very good source of carbon. The use of immobilized cells can further reduce the overall cost of the production process. This study primarily aims to produce succinic acid from glycerol residue through the use of immobilized Escherichia coli in a batch fermentation process. The parameters which affect bacterial fermentation process such as the mass substrate, temperature, inoculum size and duration of fermentation were screened using One-Factor-At-a-Time (OFAT method. The result of the screening process shows that a substrate (glycerol concentration of 30 g, inoculum size 20% v/v, and time 4 h produced the maximum succinic acid concentration of 117.99 g/L. The immobilized cells were found to be stable as well as retain their fermentative ability up to the 6th cycle of recycling, thereby presenting as an advantage over the free cell system. Therefore, conclude that using immobilized cells can contribute immensely to the cost-effective production of succinic acid from glycerol residue.

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

    International Nuclear Information System (INIS)

    1982-03-01

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

  12. The Hanford Site Tank Waste Remediation System: An update

    International Nuclear Information System (INIS)

    Alumkal, W.T.; Babad, H.; Harmon, H.D.; Wodrich, D.D.

    1994-01-01

    The U.S. Department of Energy's Hanford Site, located in southeastern Washington State, has the most diverse and largest amount of highly radioactive waste in the United States. High-level radioactive waste has been stored in large underground tanks since 1944. Approximately 230,000 m 3 (61 Mgal) of caustic liquids, slurries, saltcakes, and sludges have 137 Cs accumulated in 177 tanks. In addition, significant amounts of 90 Sr and were removed from the tank waste, converted to salts, doubly encapsulated in metal containers., and stored in water basins. A Tank Waste Remediation System Program was established by the U.S. Department of Energy in 1991 to safely manage and immobilize these wastes in anticipation of permanent disposal of the high-level waste fraction in a geologic repository. Since 1991, progress has been made resolving waste tank safety issues, upgrading Tank Farm facilities and operations, and developing a new strategy for retrieving, treating, and immobilizing the waste for disposal

  13. A summary of the program and progress to 1984 December of the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Dixon, R.S.

    1986-08-01

    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 fifth 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 1984 are described in this report. 74 refs

  14. Recovery and removal of uranium by using plant wastes

    International Nuclear Information System (INIS)

    Nakajima, Akira; Sakaguchi, Takashi

    1990-01-01

    The uranium-adsorbing abilities of seven plant wastes were investigated. High abilities to adsorb uranium from non-saline water containing 10 mg dm -3 of uranium were observed with a number of plant wastes tested. However, with seawater supplemented with 10 mg dm -3 of uranium, similar results were found only with chestnut residues. When the plant wastes were immobilized with formaldehyde, their ability to adsorb uranium was increased. Uranium and copper ions were more readily adsorbed by all plant wastes tested than other metal ions from a solution containing a mixture of seven different heavy metals. The selective adsorption of heavy metal ions differs with different species of plant wastes. The immobilization of peanut inner skin, orange peel and grapefruit peel increased the selectivity for uranium. (author)

  15. 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

  16. Description of a ceramic waste form and canister for Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Butler, J.L.; Allender, J.S.; Gould, T.H. Jr.

    1982-04-01

    A canistered ceramic waste form for possible immobilization of Savannah River Plant (SRP) high-level radioactive wastes is described. Characteristics reported for the form include waste loading, chemical composition, heat content, isotope inventory, mechanical and thermal properties, and leach rates. A conceptual design of a potential production process for making this canistered form are also described. The ceramic form was selected in November 1981 as the primary alternative to the reference waste form, borosilicate glass, for making a final waste form decision for SRP waste by FY-1983. 11 tables

  17. Economics of defense high level waste management in the United States

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1987-01-01

    Life-cycle costs of defense waste disposal, as presented in the foregoing sections, are summarized. Expressed as incremental costs per canister of waste deposited in a Federal geologic repository and per gallon of decontaminated salt solution immobilized in onsite concrete vaults, the tabulated values provide a measure of waste management costs relatively independent of the inventories of waste processed. Total values are about $350,000 per glass waste canister processed and $4.68 per gallon of decontaminated salt immobilized. These costs do not generally include contributions of fixed charges, such as capital costs, except in the case of transport and repository charges for which the quantities of waste handled determine allocation of fixed costs included in the fee assessments. 14 refs., 2 figs., 3 tabs

  18. 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

  19. Status of the Canadian Nuclear Fuel Waste Management Program

    International Nuclear Information System (INIS)

    Lyon, R.B.

    1985-10-01

    The Canadian Nuclear Fuel Waste Management Program is in the fifth year of a ten-year generic research and development phase. The major objective of this phase of the program is to assess the basic safety and environmental aspects of the concept of isolating immobilized fuel waste by deep underground disposal in plutonic rock. The major scientific and engineering components of the program, namely immobilization studies, geoscience research, and environmental and safety assessment, are well established

  20. Supported liquid inorganic membranes for nuclear waste separation

    Science.gov (United States)

    Bhave, Ramesh R; DeBusk, Melanie M; DelCul, Guillermo D; Delmau, Laetitia H; Narula, Chaitanya K

    2015-04-07

    A system and method for the extraction of americium from radioactive waste solutions. The method includes the transfer of highly oxidized americium from an acidic aqueous feed solution through an immobilized liquid membrane to an organic receiving solvent, for example tributyl phosphate. The immobilized liquid membrane includes porous support and separating layers loaded with tributyl phosphate. The extracted solution is subsequently stripped of americium and recycled at the immobilized liquid membrane as neat tributyl phosphate for the continuous extraction of americium. The sequestered americium can be used as a nuclear fuel, a nuclear fuel component or a radiation source, and the remaining constituent elements in the aqueous feed solution can be stored in glassified waste forms substantially free of americium.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  2. UTYLIZATION METHODS OF SLAGS AND ASH FROM WASTE INCINERATION PLANTS

    Directory of Open Access Journals (Sweden)

    Janusz Mikuła

    2017-06-01

    Full Text Available The paper presents modern management methods, solidification and immobilization of ash and slag from waste incineration plants. The innovative technologies for solving this kind of problem were described. Results focused on the most promising technologies of solidification, among others geopolymerization processes. The paper presents examples of the results of solidified ash and slag in the geopolymer matrix. The studies showed that the leachable of heavy metals from the geopolymer matrix containing ashes from the incineration of municipal waste qualifies them for storage in landfills for non-hazardous and inert. Moreover, these studies demonstrated practically 100% effectiveness for immobilization of the elements: bar (Ba, cadmium (Cd, zinc (Zn, mercury (Hg, nickel (Ni, lead (Pb. In the case of chromium III (Cr+3 97% level of effectiveness of the immobilization was achieved. In order to immobilize chromium VI (Cr+6 introduced additions of sulfur compounds. The study confirmed the low efficiency of the immobilization of: arsenic (As, selenium (Se and molybdenum (Mo.

  3. Systems approach to nuclear waste glass development

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1986-01-01

    Development of a host solid for the immobilization of nuclear waste has focused on various vitreous wasteforms. The systems approach requires that parameters affecting product performance and processing be considered simultaneously. Application of the systems approach indicates that borosilicate glasses are, overall, the most suitable glasses for the immobilization of nuclear waste. Phosphate glasses are highly durable; but the glass melts are highly corrosive and the glasses have poor thermal stability and low solubility for many waste components. High-silica glasses have good chemical durability, thermal stability, and mechanical stability, but the associated high melting temperatures increase volatilization of hazardous species in the waste. Borosilicate glasses are chemically durable and are stable both thermally and mechanically. The borosilicate melts are generally less corrosive than commercial glasses, and the melt temperature miimizes excessive volatility of hazardous species. Optimization of borosilicate waste glass formulations has led to their acceptance as the reference nuclear wasteform in the United States, United Kingdom, Belgium, Germany, France, Sweden, Switzerland, and Japan

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  5. 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

  6. Phase 1 Testing Results of Immobilization of WTP Effluent Management Facility Evaporator Bottoms Core Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, Alex D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-05

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator in the Effluent Management Facility (EMF) and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of this stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Alternate disposition would also eliminate this stream from recycling within WTP when it begins operations and would decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures and are problematic for the glass waste form, such as halides and sulfate, along with entrained, volatile, and semi-volatile metals, such as Hg, As, and Se. Because this stream will recycle within WTP, these components accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfate that get recycled to the melter, and is a key objective of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of earlier tasks was to formulate and prepare a

  7. 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.

  8. Ceramic waste forms for fuel-containing masses at Chernobyl

    International Nuclear Information System (INIS)

    Oversby, V.M.

    1994-05-01

    The fuel materials originally in the core of the Chernobyl Unit 4 reactor are now present within the Ukrytie in three major forms: (1) very fine particles of fuel dispersed as dust (about 10 tonnes), (2) fragments of the destroyed core, and (3) lavas containing fuel, cladding, and other materials. All of these materials will need to be immobilized into waste forms suitable for final disposal. We propose a ceramic waste form system that could accommodate all three waste types with a single set of processing equipment. The waste form would include the mineral zirconolite for immobilization of actinide materials (including uranium), perovskite, nepheline, spinel, and other phases as dictated by the chemistry of the lava masses. Waste loadings as high as 50% U can be achieved if pyrochlore, a close relative of zirconolite, is used as the U host. The ceramic immobilization could be achieved with low additions of inert chemicals to minimize the final disposal volume while ensuring a durable product. The sequence of processing would be to collect and immobilize the fuel dust first. This material will require minimal preprocessing and will provide experience in the handling of the fuel materials. Core fragments would be processed next, using a cryogenic crushing stage to reduce the size prior to adding ceramic additives. The lavas would be processed last, which is compatible with the likely sequence of availability of materials and with the complexity of the operations. The lavas will require more adjustment of chemical additive composition than the other streams to ensure that the desired phases are produced in the waste form

  9. 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

  10. Hanford Waste Vitrification Plant Technology Plan

    International Nuclear Information System (INIS)

    Sexton, R.A.

    1988-06-01

    The reference Hanford plan for disposal of defense high-level waste is based on waste immobilization in glass by the vitrification process and temporary vitrified waste storage at the Hanford Site until final disposal in a geologic repository. A companion document to the Hanford Waste Management Plan (HWMP) is the Draft, Interim Hanford Waste Management Technology Plan (HWMTP), which provides a description of the technology that must be developed to meet the reference waste management plan. One of the issues in the HWMTP is DST-6, Immobilization (Glass). The HWMTP includes all expense funding needed to complete the Hanford Waste Vitrification Plant (HWVP) project. A preliminary HWVP Technology Plan was prepared in 1985 as a supporting document to the HWMTP to provide a more detailed description of the technology needed to construct and operate a vitrification facility. The plan was updated and issued in 1986, and revised in 1987. This document is an annual update of the plan. The HWVP Technology Plan is limited in scope to technology that requires development or confirmation testing. Other expense-funded activities are not included. The relationship between the HWVP Technology Plan and other waste management issues addressed in the HWMTP is described in section 1.6 of this plan. 6 refs., 4 figs., 34 tabs

  11. WNA's Policy Document : sustaining global best practices in uranium, mining and processing, principles for managing radiation, health and safety, waste and the environment

    International Nuclear Information System (INIS)

    Saint-Pierre, S.; Waste Management and Decommissioning Working Group-WM and DW

    2008-01-01

    The worldwide community of uranium mining and processing recognizes that managing radiation, health and safety, waste and the environment is paramount. Such responsible management applies at all stages of planning and activities. Today we are acting to ensure that all parties directly involved in uranium mining and processing strive to achieve the highest levels of excellence in these fields. We are doing so by sustaining a strong safety culture based on a commitment to common, internationally shared principles. This paper sets out principles for the management of radiation, health and safety, waste and the environment applicable to sites throughout the world. In national and regional settings where nuclear fuel cycle activities are well developed, these principles already serve as the underpinning for 'Codes of Practice' that govern uranium mining and processing. In any given setting, a Code of Practice is needed to guide practical implementation of these principles according to the regional, national or site-specific context. These principles are published in the belief that they hold special relevance for emerging uranium producing countries that do not yet have fully developed regulations for the control of radiation, health and safety, waste and the environment associated with uranium mining and processing. The principles are equally relevant for operators, contractors, and regulators newly engaged in uranium mining and processing. Once national regulations are fully developed, they can be expected to embody these principles. Each principle affirmed here will not apply to the same extent for each party. Ultimately, the precise allocation of responsibilities must be set at the national and local levels. This document holds the status of a policy and ethical declaration by the full WNA membership, which the global nuclear industry. The principles affirmed here are supported by key relevant international organizations, including the IAEA and the global mining

  12. Preliminary assessment of modified borosilicate glasses for chromium and ruthenium immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Farid, Osama M. [Reactors Department, Nuclear Research Center, Atomic Energy Authority of Egypt, P.O. 13759, Inshas, Cairo (Egypt); Centre of Nuclear Engineering (CNE), Department of Materials, Imperial College London, London, SW7 2BP (United Kingdom); Abdel Rahman, R.O., E-mail: alaarehab@yahoo.com [Hot Laboratory Center, Atomic Energy Authority of Egypt, P.O. 13759, Inshas, Cairo (Egypt)

    2017-01-15

    The feasibility of using modified alkali borosilicate glasses for ruthenium and chromium immobilization is preliminary assessed by investigating the immobilization system structure under normal conditions. Within this context, reference alkali borosilicate, and simulated Magnox-modified glasses were prepared and studied. The results indicate that ruthenium is immobilized in the vitreous structure as encapsulated RuO{sub 2} crystallites that act as seeds for heterogeneous nucleation of other crystalline phases. The presence of Zn, as modifier, has contributed to chromium immobilization in zincochromite spinel structure, whereas Ca is accommodated in the vitreous structure. Immobilization performance was evaluated by conducting conservative static leach test and studying the leached glass. Leached glass morphology was altered, where near surface reference glass is leached over 400 nm and simulated Magnox-modified sample is altered over 300 nm. Normalized release rates are within normal range for borosilicate material. For simulated Magnox-modified sample, Ca and alkali structural element, i.e. Na and Li, are leached via ion-exchange reaction. The ion-exchanged fraction equals 1.06 × 10{sup −8} mol/m{sup 2} s and chromium has slightly lower normalized release rate value than ruthenium. - Highlights: • The presence of modifiers and waste oxides led to localized de-vitrification. • Ruthenium is encapsulated within the vitreous glass network as RuO{sub 2} crystals. • Chromium is immobilized within the zincochromite spinel structure. • Pitting and cracks induced by leaching did not affect the immobilization performance.

  13. Low-level waste program technical strategy

    International Nuclear Information System (INIS)

    Bledsoe, K.W.

    1994-01-01

    The Low-Level Waste Technical Strategy document describes the mechanisms which the Low-Level Waste Program Office plans to implement to achieve its mission. The mission is to manage the receipt, immobilization, packaging, storage/disposal and RCRA closure (of the site) of the low-level Hanford waste (pretreated tank wastes) in an environmentally sound, safe and cost-effective manner. The primary objective of the TWRS Low-level waste Program office is to vitrify the LLW fraction of the tank waste and dispose of it onsite

  14. 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...

  15. 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

  16. Investigation of vitreous and crystalline ceramic materials for immobilization of alpha-contaminated residues

    International Nuclear Information System (INIS)

    Palmer, C.R.; Mellinger, G.B.; Rusin, J.M.

    1981-01-01

    Experimental investigations of two alternatives for immobilizing dispersible solid wastes contaminated with alpha-emitting radionuclides are reviewed. Borosilicate glasses and sintered silicate ceramics are being studied for such wastes, and results so far indicate both may offer attractive alternatives to waste generators. Waste oxide solubilities, de-vitrification behaviour and effects of residual carbon are examined for glasses incorporating incinerator ash and hydrated ferric oxide sludge. Glasses will accommodate these wastes at loadings of 30-60 wt% while maintaining good performance characteristics. A brief comparative evaluation of cold-pressed and sintered ceramics is also described. The effects on process and product properties of the choice of additives, waste loading and sintering temperature were determined. This approach also appears to promise economic waste loadings while achieving relatively durable waste forms. (author)

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

    International Nuclear Information System (INIS)

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

    1993-10-01

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

  18. Division of Waste Management, Production, and Reprocessing programs progress report for January--December 1976

    International Nuclear Information System (INIS)

    Lerch, R.E.

    1977-04-01

    Development of the acid digestion process for treating combustible nuclear wastes has progressed to design and construction of the Radioactive Acid Digestion Test Unit (RADTU). Tests were continued in the nonradioactive Acid Digestion Test Unit (ADTU) in an effort to improve the performance of the system. Nitric acid consumption has been decreased from 8.8 to 4.5 kg HNO 3 /kg digested waste by adding the nitric acid deep below the liquid surface in the annular heating vessel instead of at the surface of the tray digester. A highly successful 70-hr continuous processing run was completed in the ADTU to confirm the operating experience in shorter runs and to test the use of air (instead of nitrogen) in the airlift pump. Initial studies were completed on fixation of acid digestion residue following recovery and drying. A variety of solidification and testing equipment was assembled to prepare and test immobilized waste products. Studies were continued on immobilization of salts and wet wastes. Salt residue-cement systems characterized include calcium sulfate, ferric sulfate, sodium sulfate, sodium nitrate, calcium chloride, and sodium chloride. Anion and cation exchange resin immobilization in cement has also been studied, as well as immobilization of sodium silicate in cement. A new program on Intermediate Level Liquid Waste (ILLW) Solidification was started with the purpose of developing and demonstrating immobilization technologies for liquid and particulate solid ILW within the fuel cycle. The four primary fuel cycle operations (reactors, spent fuel storage basins, fuel reprocessing plants, and MOX fuel fabrication plants) were reviewed and a list of sources and quantities of waste was formulated. The annual accumulation of radioactive waste materials at the six commercial waste burial sites was updated through 1976. The total accumulated volume at the end of 1976 was about 434,000 m 3

  19. Solidification of TRU wastes in a ceramic matrix

    International Nuclear Information System (INIS)

    Loida, A.; Schubert, G.

    1991-01-01

    Aluminumsilicate based ceramic materials have been evaluated as an alternative waste form for the incorporation of TRU wastes. These waste forms are free of water and - cannot generate hydrogen radiolyticly, - they show good compatibility between the compounds of the waste and the matrix, - they are resistent against aqueous solutions, heat and radiation. R and D-work has been performed to demonstrate the suitability of this waste form for the immobilization of TRU-wastes. Four kinds of original TRU-waste streams and a mixture of all of them have been immobilized by ceramization, using glove box and remote operation technique as well. Clay minerals, (kaolinite, bentonite) and reactive corundum were selected as ceramic raw materials (KAB 78) in an appropriate ratio yielding 78 wt% Al 2 O 3 and 22 wt%SiO 2 . The main process steps are (i) pretreatment of the liquid waste (concentration, denitration, neutralization, solid- liquid separation), (ii) mixing with ceramic raw materials and forming, (iii) heat treatment with T max. of 1300 0 C for 15 minutes. The waste load of the ceramic matrix has been increased gradually from 20 to 50, in some cases to 60 wt.%

  20. 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

  1. 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.

  2. The defense waste processing facility: the final processing step for defense high-level waste disposal

    International Nuclear Information System (INIS)

    Cowan, S.P.; Sprecher, W.M.; Walton, R.D.

    1983-01-01

    The policy of the U.S. Department of Energy is to pursue an aggressive and credible waste management program that advocates final disposal of government generated (defense) high-level nuclear wastes in a manner consistent with environmental, health, and safety responsibilities and requirements. The Defense Waste Processing Facility (DWPF) is an essential component of the Department's program. It is the first project undertaken in the United States to immobilize government generated high-level nuclear wastes for geologic disposal. The DWPF will be built at the Department's Savannah River Plant near Aiken, South Carolina. When construction is complete in 1989, the DWPF will begin processing the high-level waste at the Savannah River Plant into a borosilicate glass form, a highly insoluble and non-dispersable product, in easily handled canisters. The immobilized waste will be stored on site followed by transportation to and disposal in a Federal repository. The focus of this paper is on the DWPF. The paper discusses issues which justify the project, summarizes its technical attributes, analyzes relevant environmental and insitutional factors, describes the management approach followed in transforming technical and other concepts into concrete and steel, and concludes with observations about the future role of the facility

  3. The DWPF waste form qualification program

    International Nuclear Information System (INIS)

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

    1994-01-01

    Prior to the introduction of radioactive feed into the Defense Waste Processing Facility for immobilization in borosilicate glass an extensive waste qualification program must be completed. The DWPF must demonstrate its ability to comply with the Waste Acceptance Product Specifications. This ability is being demonstrated through laboratory and pilot scale work and will be completed after the full operation of the DWPF using various simulated feeds

  4. Transportation and disposal of low-and medium level waste using fiber reinforced concrete overpacks

    International Nuclear Information System (INIS)

    Pech, R.; Verdier, A.

    1993-01-01

    A multiple-year research effort by Cogema culminated in the development of a new process to immobilize nuclear waste in concrete overpacks reinforced with metal fibers. The fiber concrete overpacks satisfy all French safety requirements relating to waste immobilization and disposal, and have been certified by Andra, the national radioactive waste management agency. This presentation will cover the use of the fiber-reinforced concrete overpack for disposal and transportation, and will discuss their fabrication. (J.P.N.)

  5. Molten salt processing of mixed wastes with offgas condensation

    International Nuclear Information System (INIS)

    Cooper, J.F.; Brummond, W.; Celeste, J.; Farmer, J.; Hoenig, C.; Krikorian, O.H.; Upadhye, R.; Gay, R.L.; Stewart, A.; Yosim, S.

    1991-01-01

    We are developing an advanced process for treatment of mixed wastes in molten salt media at temperatures of 700--1000 degrees C. Waste destruction has been demonstrated in a single stage oxidation process, with destruction efficiencies above 99.9999% for many waste categories. The molten salt provides a heat transfer medium, prevents thermal surges, and functions as an in situ scrubber to transform the acid-gas forming components of the waste into neutral salts and immobilizes potentially fugitive materials by a combination of particle wetting, encapsulation and chemical dissolution and solvation. Because the offgas is collected and assayed before release, and wastes containing toxic and radioactive materials are treated while immobilized in a condensed phase, the process avoids the problems sometimes associated with incineration processes. We are studying a potentially improved modification of this process, which treats oxidizable wastes in two stages: pyrolysis followed by catalyzed molten salt oxidation of the pyrolysis gases at ca. 700 degrees C. 15 refs., 5 figs., 1 tab

  6. Institute for Nuclear Waste Disposal. Annual Report 2011

    International Nuclear Information System (INIS)

    Geckeis, H.; Stumpf, T.

    2012-01-01

    The R and D at the Institute for Nuclear Waste Disposal, INE, (Institut fuer Nukleare Entsorgung) of the Karlsruhe Institute of Technology (KIT) focuses on (i) long term safety research for nuclear waste disposal, (ii) immobilization of high level radioactive waste (HLW), (iii) separation of minor actinides from HLW and (iv) radiation protection.

  7. Challenges in development of matrices for vitrification of old legacy waste and high-level radioactive waste generated from reprocessing of AHWR and FBR spent fuel

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2012-01-01

    Majority of radioactivity in entire nuclear fuel cycle is concentrated in HLW. A three step strategy for management of HLW has been adopted in India. This involves immobilization of waste oxides in stable and inert solid matrices, interim retrievable storage of the conditioned waste product under continuous cooling and disposal in deep geological formations. Glass has been accepted as most suitable matrix world-wide for immobilization of HLW, because of its attractive features like ability to accommodate wide range of waste constituents, modest processing temperatures, adequate chemical, thermal and radiation stability. Borosilicate glass matrix developed by BARC in collaboration with CGCRI has been adopted in India for immobilization of HLW. In view of compositional variation of HLW from site to site, tailor make changes in the glass formulations are often necessary to incorporate all the waste constituents and having the product of desirable characteristics. The vitrified waste products made with different glass formulations and simulated waste need to be characterized for chemical durability, thermal stability, homogeneity etc. before finalizing a suitable glass formulation. The present extended abstract summarises the studies carried out for development of glass formulations for vitrification of legacy waste and futuristic waste likely to be generated from AHWR and FBR having wide variations in their compositions. The presently stored HLW at Trombay is characterized by significant concentrations of uranium, sodium and sulphate in addition to fission products, corrosion products and small amount of other actinides

  8. Secondary Waste Form Down Selection Data Package – Ceramicrete

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Westsik, Joseph H.

    2011-08-31

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

  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. 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

  11. Disposal criticality analysis methodology for fissile waste forms

    International Nuclear Information System (INIS)

    Davis, J.W.; Gottlieb, P.

    1998-03-01

    A general methodology has been developed to evaluate the criticality potential of the wide range of waste forms planned for geologic disposal. The range of waste forms include commercial spent fuel, high level waste, DOE spent fuel (including highly enriched), MOX using weapons grade plutonium, and immobilized plutonium. The disposal of these waste forms will be in a container with sufficiently thick corrosion resistant barriers to prevent water penetration for up to 10,000 years. The criticality control for DOE spent fuel is primarily provided by neutron absorber material incorporated into the basket holding the individual assemblies. For the immobilized plutonium, the neutron absorber material is incorporated into the waste form itself. The disposal criticality analysis methodology includes the analysis of geochemical and physical processes that can breach the waste package and affect the waste forms within. The basic purpose of the methodology is to guide the criticality control features of the waste package design, and to demonstrate that the final design meets the criticality control licensing requirements. The methodology can also be extended to the analysis of criticality consequences (primarily increased radionuclide inventory), which will support the total performance assessment for the respository

  12. 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

  13. 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

  14. 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...

  15. Tank Waste Remediation System decisions and risk assessment

    International Nuclear Information System (INIS)

    Johnson, M.E.

    1994-09-01

    The Tank Waste Remediation System (TWRS) mission is to store, treat, and immobilize the highly radioactive Hanford Site tank wastes and encapsulated cesium and strontium materials in an environmentally sound, safe, and cost effective manner. Additionally, the TWRS conducts, as part of this mission, resolution of safety issues associated with the wastes within the 177 underground radioactive waste tanks. Systems engineering principles are being applied to determine the functions and establish requirements necessary for accomplishing the TWRS mission (DOE 1994 draft). This systematic evaluation of the TWRS program has identified key decisions that must be executed to establish mission scope, determine requirements, or select a technical solution for accomplishing identified functions and requirements. Key decisions identified through the systematic evaluation of the TWRS mission are presented in this document. Potential alternative solutions to each decision are discussed. After-discussion and evaluation of each decision with effected stakeholder groups, the US Department of Energy (DOE) will select a solution from the identified alternatives for implementation. In order to proceed with the development and execution of the tank waste remediation program, the DOE has adopted a planning basis for several of these decisions, until a formal basis is established. The planning bases adopted by the DOE is continuing to be discussed with stakeholder groups to establish consensus for proceeding with proposed actions. Technical and programmatic risks associated with the planning basis adopted by the DOE are discussed

  16. Life cycle costing of waste management systems: overview, calculation principles and case studies.

    Science.gov (United States)

    Martinez-Sanchez, Veronica; Kromann, Mikkel A; Astrup, Thomas Fruergaard

    2015-02-01

    This paper provides a detailed and comprehensive cost model for the economic assessment of solid waste management systems. The model was based on the principles of Life Cycle Costing (LCC) and followed a bottom-up calculation approach providing detailed cost items for all key technologies within modern waste systems. All technologies were defined per tonne of waste input, and each cost item within a technology was characterised by both a technical and an economic parameter (for example amount and cost of fuel related to waste collection), to ensure transparency, applicability and reproducibility. Cost items were classified as: (1) budget costs, (2) transfers (for example taxes, subsidies and fees) and (3) externality costs (for example damage or abatement costs related to emissions and disamenities). Technology costs were obtained as the sum of all cost items (of the same type) within a specific technology, while scenario costs were the sum of all technologies involved in a scenario. The cost model allows for the completion of three types of LCC: a Conventional LCC, for the assessment of financial costs, an Environmental LCC, for the assessment of financial costs whose results are complemented by a Life Cycle Assessment (LCA) for the same system, and a Societal LCC, for socio-economic assessments. Conventional and Environmental LCCs includes budget costs and transfers, while Societal LCCs includes budget and externality costs. Critical aspects were found in the existing literature regarding the cost assessment of waste management, namely system boundary equivalency, accounting for temporally distributed emissions and impacts, inclusions of transfers, the internalisation of environmental impacts and the coverage of shadow prices, and there was also significant confusion regarding terminology. The presented cost model was implemented in two case study scenarios assessing the costs involved in the source segregation of organic waste from 100,000 Danish households and

  17. 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...

  18. Study on immobilization technology of krypton gas by ion-implantation and sputtering process

    International Nuclear Information System (INIS)

    Hayashi, S.; Kamiya, S.; Ikeda, S.; Nakanishi, Y.

    1997-01-01

    At the Waste Technology Development Division of PNC''s Tokai Works, we have been carrying out the development of the recovering and storing technology of radioactive krypton to reduce the radioactive gaseous effluent released to the environment arising from reprocessing of nuclear spent fuels. As a part of this technical development, the method of storing radioactive krypton is an encapsulation in a metal matrix which appears to offer the most secure immobilization of krypton. From the result of the ion-implantation test used non-radioactive krypton and the evaluation of electrical characteristics, we have selected the metallic materials, they are Nickel and Yttrium, and designed the vessel and the apparatus which continuously immobilized the krypton into metal matrix. This article described the following experiment result. (1) The long-term immobilization technology was established with respect to handling a equipment and plasma controlling. (2) Stability of the krypton immobilized alloy was considered satisfactory because the release was tolerably small. (author)

  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. 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