WorldWideScience

Sample records for waste forms

  1. Densified waste form and method for forming

    Science.gov (United States)

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

    2015-08-25

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

  2. Comparative waste forms study

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-12-01

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

  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. CERAMIC WASTE FORM DATA PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J.; Marra, J.

    2014-06-13

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

  5. Low temperature waste form process intensification

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Randklev, E.H.

    1993-06-01

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

  7. 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

  8. Alternative solidified forms for nuclear wastes

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-01-01

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

  9. 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.

  10. Korean Waste Management Law and Waste Disposal Forms.

    Science.gov (United States)

    1991-03-01

    Soil Treatment Tanks) 69 Article 8. (Interim Measures on Report of Recycler or Reuser of Industrial Waste) 69 Article 9. (Interim Measures on Permit...recycling and reuse (hereinafter referred to as a "recycler and reuser of industrial waste"), pursuant to Article 23.2. of the Law, shall submit a "Filing... reuser of industrial waste, pursuant to Article 45.2., shall submit a "Modification of Recycle or Reuse of Industrial Waste" (Form No. 17), to the

  11. Combined Waste Form Cost Trade Study

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-11-01

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

  12. Miscellaneous Waste-Form FEPs

    Energy Technology Data Exchange (ETDEWEB)

    A. Schenker

    2000-12-08

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

  13. Radionuclide Retention in Concrete Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-30

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

  14. 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.

  15. Ceramic and glass radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-01-01

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

  16. Performance Test on Polymer Waste Form

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  17. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

  1. CRYSTALLINE CERAMIC WASTE FORMS: REFERENCE FORMULATION REPORT

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

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

  2. Liquid Secondary Waste Grout Formulation and Waste Form Qualification

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-23

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

  3. SEPARATIONS AND WASTE FORMS CAMPAIGN IMPLEMENTATION PLAN

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-26

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

  4. Development of Alternative Technetium Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Czerwinski, Kenneth

    2013-09-13

    The UREX+1 process is under consideration for the separation of transuranic elements from spent nuclear fuel. The first steps of this process extract the fission product technicium-99 ({sup 99}Tc) into an organic phase containing tributylphosphate together with uranium. Treatment of this stream requires the separation of Tc from U and placement into a suitable waste storage form. A potential candidate waste form involves immobilizing the Tc as an alloy with either excess metallic zirconium or stainless steel. Although Tc-Zr alloys seem to be promising waste forms, alternative materials must be investigated. Innovative studies related to the synthesis and behavior of a different class of Tc materials will increase the scientific knowledge related to development of Tc waste forms. These studies will also provide a better understanding of the behavior of {sup 99}Tc in repository conditions. A literature survey has selected promising alternative waste forms for further study: technetium metallic alloys, nitrides, oxides, sulfides, and pertechnetate salts. The goals of this project are to 1) synthesize and structurally characterize relevant technetium materials that may be considered as waste forms, 2) investigate material behavior in solution under different conditions of temperature, electrochemical potential, and radiation, and 3) predict the long-term behavior of these materials.

  5. IGNEOUS INTRUSION IMPACTS ON WASTE PACKAGES AND WASTE FORMS

    Energy Technology Data Exchange (ETDEWEB)

    P. Bernot

    2004-04-19

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

  6. Alternative Waste Forms for Electro-Chemical Salt Waste

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-28

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    Science.gov (United States)

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

    2014-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-15

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

  11. Alternative High-Performance Ceramic Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Sundaram, S. K. [Alfred Univ., NY (United States)

    2017-02-01

    This final report (M5NU-12-NY-AU # 0202-0410) summarizes the results of the project titled “Alternative High-Performance Ceramic Waste Forms,” funded in FY12 by the Nuclear Energy University Program (NEUP Project # 12-3809) being led by Alfred University in collaboration with Savannah River National Laboratory (SRNL). The overall focus of the project is to advance fundamental understanding of crystalline ceramic waste forms and to demonstrate their viability as alternative waste forms to borosilicate glasses. We processed single- and multiphase hollandite waste forms based on simulated waste streams compositions provided by SRNL based on the advanced fuel cycle initiative (AFCI) aqueous separation process developed in the Fuel Cycle Research and Development (FCR&D). For multiphase simulated waste forms, oxide and carbonate precursors were mixed together via ball milling with deionized water using zirconia media in a polyethylene jar for 2 h. The slurry was dried overnight and then separated from the media. The blended powders were then subjected to melting or spark plasma sintering (SPS) processes. Microstructural evolution and phase assemblages of these samples were studied using x-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersion analysis of x-rays (EDAX), wavelength dispersive spectrometry (WDS), transmission electron spectroscopy (TEM), selective area x-ray diffraction (SAXD), and electron backscatter diffraction (EBSD). These results showed that the processing methods have significant effect on the microstructure and thus the performance of these waste forms. The Ce substitution into zirconolite and pyrochlore materials was investigated using a combination of experimental (in situ XRD and x-ray absorption near edge structure (XANES)) and modeling techniques to study these single phases independently. In zirconolite materials, a transition from the 2M to the 4M polymorph was observed with increasing Ce content. The resulting

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-21

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

  13. Accelerated leach test of paraffin waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Kim, Ju Youl; Cho, Young Ok; Park, Joo Wan [Nuclear Environment Technology Institute, KEPCO, Taejon (Korea, Republic of)

    1999-07-01

    Leach test for the paraffin waste forms, which was recently introduced to immobilize the dry radioactive waste concentrates at the nuclear power plants in Korea, was conducted in accordance with the Accelerated Leach Test adopted by ASTM as Standard Test Method. The specimens were made of 22 w% paraffin, 78 w% boric acid, and little amount of Co, Sr. and Cs to determine the leaching mechanism. Leach tests for the specimens with different amounts of additives were conducted at three temperatures to investigate the effect of additives on the leach rates. The leach rates of boric acid, Co, Sr, and Cs were measured. The results showed that boric acid and Co leached congruently. The leaching rates are dependent on the temperature as expected. The ALT computer program was used to simulate the experimental data. The ALT program calculation shows that the diffusion can not explain the experimental data. (author)

  14. Electrochemical corrosion testing of metal waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Abraham, D. P.; Peterson, J. J.; Katyal, H. K.; Keiser, D. D.; Hilton, B. A.

    1999-12-14

    Electrochemical corrosion tests have been conducted on simulated stainless steel-zirconium (SS-Zr) metal waste form (MWF) samples. The uniform aqueous corrosion behavior of the samples in various test solutions was measured by the polarization resistance technique. The data show that the MWF corrosion rates are very low in groundwaters representative of the proposed Yucca Mountain repository. Galvanic corrosion measurements were also conducted on MWF samples that were coupled to an alloy that has been proposed for the inner lining of the high-level nuclear waste container. The experiments show that the steady-state galvanic corrosion currents are small. Galvanic corrosion will, hence, not be an important mechanism of radionuclide release from the MWF alloys.

  15. Monazite as a suitable actinide waste form

    Energy Technology Data Exchange (ETDEWEB)

    Schlenz, Hartmut; Heuser, Julia; Schmitz, Stephan; Bosbach, Dirk [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Energie und Klimaforschung (IEK), Nukleare Entsorgung und Reaktorsicherheit (IEK-6); Neumann, Andreas [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Energie und Klimaforschung (IEK), Nukleare Entsorgung und Reaktorsicherheit (IEK-6); RWTH Aachen Univ. (Germany). Inst. for Crystallography

    2013-03-01

    The conditioning of radioactive waste from nuclear power plants and in some countries even of weapons plutonium is an important issue for science and society. Therefore the research on appropriate matrices for the immobilization of fission products and actinides is of great interest. Beyond the widely used borosilicate glasses, ceramics are promising materials for the conditioning of actinides like U, Np, Pu, Am, and Cm. Monazite-type ceramics with general composition LnPO{sub 4} (Ln = La to Gd) and solid solutions of monazite with cheralite or huttonite represent important materials in this field. Monazite appears to be a promising candidate material, especially because of its outstanding properties regarding radiation resistance and chemical durability. This article summarizes the most recent results concerning the characterization of monazite and respective solid solutions and the study of their chemical, thermal, physical and structural properties. The aim is to demonstrate the suitability of monazite as a secure and reliable waste form for actinides. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-01

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

  17. Safeguards and retrievability from waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Danker, W.

    1996-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Rusin, J.M.

    1980-12-01

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

  20. Equilibrium Temperature Profiles within Fission Product Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Kaminski, Michael D. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-10-01

    We studied waste form strategies for advanced fuel cycle schemes. Several options were considered for three waste streams with the following fission products: cesium and strontium, transition metals, and lanthanides. These three waste streams may be combined or disposed separately. The decay of several isotopes will generate heat that must be accommodated by the waste form, and this heat will affect the waste loadings. To help make an informed decision on the best option, we present computational data on the equilibrium temperature of glass waste forms containing a combination of these three streams.

  1. CSNF WASTE FORM DEGRADATION: SUMMARY ABSTRACTION

    Energy Technology Data Exchange (ETDEWEB)

    J.C. CUNNANE

    2004-08-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  3. Effluent Management Facility Evaporator Bottom-Waste Streams Formulation and Waste Form Qualification Testing

    Energy Technology Data Exchange (ETDEWEB)

    Saslow, Sarah A.; Um, Wooyong; Russell, Renee L.

    2017-08-02

    This report describes the results from grout formulation and cementitious waste form qualification testing performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). These results are part of a screening test that investigates three grout formulations proposed for wide-range treatment of different waste stream compositions expected for the Hanford Effluent Management Facility (EMF) evaporator bottom waste. This work supports the technical development need for alternative disposition paths for the EMF evaporator bottom wastes and future direct feed low-activity waste (DFLAW) operations at the Hanford Site. High-priority activities included simulant production, grout formulation, and cementitious waste form qualification testing. The work contained within this report relates to waste form development and testing, and does not directly support the 2017 Integrated Disposal Facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY 2017 and future waste form development efforts. The provided results and data should be used by (1) cementitious waste form scientists to further the understanding of cementitious leach behavior of contaminants of concern (COCs), (2) decision makers interested in off-site waste form disposal, and (3) the U.S. Department of Energy, their Hanford Site contractors and stakeholders as they assess the IDF PA program at the Hanford Site. The results reported help fill existing data gaps, support final selection of a cementitious waste form for the EMF evaporator bottom waste, and improve the technical defensibility of long-term waste form risk estimates.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-21

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

  5. Final report on cermet high-level waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Kobisk, E.H.; Quinby, T.C.; Aaron, W.S.

    1981-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-22

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

  7. Quality control of cemented waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Slate, L.J.

    1994-12-31

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

  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. Minerals as natural analogues for crystalline nuclear waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  10. Secondary Waste Cementitious Waste Form Data Package for the Integrated Disposal Facility Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Serne, R Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-16

    A review of the most up-to-date and relevant data currently available was conducted to develop a set of recommended values for use in the Integrated Disposal Facility (IDF) performance assessment (PA) to model contaminant release from a cementitious waste form for aqueous wastes treated at the Hanford Effluent Treatment Facility (ETF). This data package relies primarily upon recent data collected on Cast Stone formulations fabricated with simulants of low-activity waste (LAW) and liquid secondary wastes expected to be produced at Hanford. These data were supplemented, when necessary, with data developed for saltstone (a similar grout waste form used at the Savannah River Site). Work is currently underway to collect data on cementitious waste forms that are similar to Cast Stone and saltstone but are tailored to the characteristics of ETF-treated liquid secondary wastes. Recommended values for key parameters to conduct PA modeling of contaminant release from ETF-treated liquid waste are provided.

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

    Science.gov (United States)

    Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

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

  13. Challenges in Modeling the Degradation of Ceramic Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Gao, Fei; Sun, Xin

    2011-09-01

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

  14. Challenges in Modeling the Degradation of Ceramic Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Gao, Fei; Sun, Xin

    2011-09-01

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

  15. Weathering Effects on Technetium Leachability from Ceramicrete Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jong Kwon; Fadzil, Syazwani Mohd; Um, Woo Yong [Pohang University of Science and Technology, Pohang (Korea, Republic of)

    2012-05-15

    Ceramicrete waste form was developed as part of the U.S. Department of Energy's (DOE's) Office of Environmental Management program to stabilize and contain volatile radioactive contaminant such as technetium ({sup 99}Tc). Ceramicrete processing technology has been demonstrated on various waste streams and has shown to retain both radioactive and hazardous contaminants effectively. Radioactive Tc is highly soluble and mobile in the environment as pertechnetate anion. Tc is also easily volatilized. Tc can be in the waste in two oxidation states . IV and VII. Tc is volatilized even during the evaporation stage when removing excess water from HLW that contains acidic solutions of Tc(VII) as pertechnetate (TcO{sub 4}){sup -}. Common chemical weathering process to occur within waste forms in the nuclear waste repository is carbonation. In addition, since technetium ({sup 99}Tc) leachability is closely related with oxidation condition and the oxidized Tc species, pertechnetate (TcO{sub 4}{sup -}) shows much higher leachability, oxidative weathering pre-treatment of waste form is important, especially for the shallow-depth radioactive waste repositories. In 2011, an evaluation of weathering effects on Tc release from different waste forms (Cast Stone and DuraLith) was conducted in the environmental chamber with different gas mixtures to produce enhanced oxidizing or carbonation conditions. Based on the technical literature and previous testing results, Ceramicrete waste form was also selected for further weathering testing to evaluate oxidizing or carbonation effects on Tc release after weathering. Leachability Indexes (LI) of Tc from two waste forms (Cast Stone and DuraLith) without pre-treatment of O{sub 2}(g) or CO{sub 2}(g) are higher than those of waste forms with pre-treatment of the same gases to simulate enhanced oxidation and carbonation weathering conditions, respectively. The LI values of two waste forms with and without weathering are shown in Table 1

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-03-01

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

  17. DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

    Energy Technology Data Exchange (ETDEWEB)

    J. CUNNANE

    2004-11-19

    Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an &apos

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-08-12

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

  19. Updated Liquid Secondary Waste Grout Formulation and Preliminary Waste Form Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Saslow, Sarah A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Russell, Renee L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Asmussen, Robert M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sahajpal, Rahul [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-07-01

    This report describes the results from liquid secondary waste grout (LSWG) formulation and cementitious waste form qualification tests performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). New formulations for preparing a cementitious waste form from a high-sulfate liquid secondary waste stream simulant, developed for Effluent Management Facility (EMF) process condensates merged with low activity waste (LAW) caustic scrubber, and the release of key constituents (e.g. 99Tc and 129I) from these monoliths were evaluated. This work supports a technology development program to address the technology needs for Hanford Site Effluent Treatment Facility (ETF) liquid secondary waste (LSW) solidification and supports future Direct Feed Low-Activity Waste (DFLAW) operations. High-priority activities included simulant development, LSWG formulation, and waste form qualification. The work contained within this report relates to waste form development and testing and does not directly support the 2017 integrated disposal facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY17, and for future waste form development efforts. The provided data should be used by (i) cementitious waste form scientists to further understanding of cementitious dissolution behavior, (ii) IDF PA modelers who use quantified constituent leachability, effective diffusivity, and partitioning coefficients to advance PA modeling efforts, and (iii) the U.S. Department of Energy (DOE) contractors and decision makers as they assess the IDF PA program. The results obtained help fill existing data gaps, support final selection of a LSWG waste form, and improve the technical defensibility of long-term waste form performance estimates.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-09-01

    This report provides a summary of the work conducted for the Waste Form Development/Test Program at Brookhaven National Laboratory in FY 1983 under the sponsorship of the US Department of Energy's Low-Level Waste Management Program. The primary focus of this work is the investigation of new solidification agents which will provide improved immobilization of low-level radioactive wastes in an efficient, cost-effective manner. A working set of preliminary waste form evaluation criteria which could impact upon the movement of radionuclides in the disposal environment was developed. The selection of potential solidification agents for further investigation is described. Two thermoplastic materials, low-density polyethylene and a modified sulfur cement were chosen as primary candidates for further study. Three waste types were selected for solidification process development and waste form property evaluation studies which represent both new volume reduction wastes (dried evaporator concentrates and incinerator ash) and current problem wastes (ion exchange resins). Preliminary process development scoping studies were conducted to verify the compatibility of selected solidification agents and waste types and the potential for improved solidification. Waste loadings of 60 wt % Na/sub 2/SO/sub 4/, 25 wt % H/sub 3/BO/sub 3/, 25 wt % incinerator ash and 50 wt % dry ion exchange resin were achieved using low density polyethylene as a matrix material. Samples incorporating 65 wt % Na/sub 2/SO/sub 4/, 40 wt % H/sub 3/BO/sub 3/, 20 wt % incinerator ash and 40 wt % dry ion exchange resin were successfully solidified in modified sulfur cement. Additional improvements are expected for both matrix materials as process parameters are optimized. Several preliminary property evaluation studies were performed to provide the basis for an initial assessment of waste form acceptability. These included a two-week water immersion test and compressive load testing.

  1. Transuranic contaminated waste form characterization and data base

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-23

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

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

    Science.gov (United States)

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

    2014-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  5. Effect of Concrete Waste Form Properties on Radionuclide Migration

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-09-30

    Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation) the mechanism of contaminant release, the significance of contaminant release pathways, how waste form performance is affected by the full range of environmental conditions within the disposal facility, the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility, the effect of waste form aging on chemical, physical, and radiological properties and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the waste forms come in contact with groundwater. Numerous sets of tests were initiated in fiscal years (FY) 2006-2009 to evaluate (1) diffusion of iodine (I) and technetium (Tc) from concrete into uncontaminated soil after 1 and 2 years, (2) I and rhenium (Re) diffusion from contaminated soil into fractured concrete, (3) I and Re (set 1) and Tc (set 2) diffusion from fractured concrete into uncontaminated soil, (4) evaluate the moisture distribution profile within the sediment half-cell, (5) the reactivity and speciation of uranium (VI) (U(VI)) compounds in concrete porewaters, (6) the rate of dissolution of concrete monoliths, and (7) the diffusion of simulated tank waste into concrete.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-02-10

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

  7. 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.

  8. Multibarrier waste forms. Part II. Characterization and evaluation.

    Energy Technology Data Exchange (ETDEWEB)

    Rusin, J.M.; Gray, W.J.; Wald, J.W.

    1979-08-01

    The multibarrier concept for the storage of radioactive waste is to use up to three barriers to isolate radionuclides from the environment: a solidified waste inner core, an impervious coating, and a metal matrix. The four multibarrier waste forms were evaluated for thermal stability (volatility), mechanical strength (impact resistance), and leach resistance. This report discusses the characterization of the multibarrier waste forms and compares them to reference calcine and glass waste forms. The weight loss of supercalcine-ceramics after 4 h in dry air ranges between 0.01 and 1.6 wt % from 1000 to 1200/sup 0/C and is dependent upon composition. Glass marbles in a cast lead alloy offer approximately an order of magnitude decease in the wt % fines < 37 ..mu..m released after impact as compared to a glass monolith. CVD-coated supercalcine in a sintered 410 SS matrix offers up to two orders of magnitude decrease. Hot-pressed supercalcine ceramics may offer no increase in impact resistance or leach resistance over that of a glass monolith. Supercalcine may offer no advantage over waste glasses in leach resistance. Glass and PyC/Al/sub 2/O/sub 3/ coatings provide effective inert leaching barriers.

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

    Energy Technology Data Exchange (ETDEWEB)

    Alexandra Navrotsky

    2010-07-30

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

  10. Technetium Waste Form Development - Progress Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-07

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  13. Reference Alloy Waste Form Fabrication and Initiation of Reducing Atmosphere and Reductive Additives Study on Alloy Waste Form Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    S.M. Frank; T.P. O' Holleran; P.A. Hahn

    2011-09-01

    This report describes the fabrication of two reference alloy waste forms, RAW-1(Re) and RAW-(Tc) using an optimized loading and heating method. The composition of the alloy materials was based on a generalized formulation to process various proposed feed streams resulting from the processing of used fuel. Waste elements are introduced into molten steel during alloy fabrication and, upon solidification, become incorporated into durable iron-based intermetallic phases of the alloy waste form. The first alloy ingot contained surrogate (non-radioactive), transition-metal fission products with rhenium acting as a surrogate for technetium. The second alloy ingot contained the same components as the first ingot, but included radioactive Tc-99 instead of rhenium. Understanding technetium behavior in the waste form is of particular importance due the longevity of Tc-99 and its mobility in the biosphere in the oxide form. RAW-1(Re) and RAW-1(Tc) are currently being used as test specimens in the comprehensive testing program investigating the corrosion and radionuclide release mechanisms of the representative alloy waste form. Also described in this report is the experimental plan to study the effects of reducing atmospheres and reducing additives to the alloy material during fabrication in an attempt to maximize the oxide content of waste streams that can be accommodated in the alloy waste form. Activities described in the experimental plan will be performed in FY12. The first aspect of the experimental plan is to study oxide formation on the alloy by introducing O2 impurities in the melt cover gas or from added oxide impurities in the feed materials. Reducing atmospheres will then be introduced to the melt cover gas in an attempt to minimize oxide formation during alloy fabrication. The second phase of the experimental plan is to investigate melting parameters associated with alloy fabrication to allow the separation of slag and alloy components of the melt.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-08-01

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

  16. Advanced waste forms research and development. Annual report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, G.J.

    1975-06-11

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

  17. Degradation modeling of the ANL ceramic waste form

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H.; Morss, L. R.

    2000-03-28

    A ceramic waste form composed of glass-bonded sodalite is being developed at Argonne National Laboratory (ANL) for immobilization and disposition of the molten salt waste stream from the electrometallurgical treatment process for metallic DOE spent nuclear fuel. As part of the spent fuel treatment program at ANL, a model is being developed to predict the long-term release of radionuclides under repository conditions. Dissolution tests using dilute, pH-buffered solutions have been conducted at 40, 70, and 90 C to determine the temperature and pH dependence of the dissolution rate. Parameter values measured in these tests have been incorporated into the model, and preliminary repository performance assessment modeling has been completed. Results indicate that the ceramic waste form should be acceptable in a repository environment.

  18. Synthesis of apatite and monazite waste form for immobilization of rare earth oxide radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, B. G.; Park, H. S.; Kim, I. T.; Lee, H. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-08-15

    In order to fabricate a monolithic waste form containing RE oxides, a vitrification at a high temperature or a ceramization by a HIP method is required. In this study, a series of monolithic wasteform with high waste loading were successfully produced at a mild condition, where the chemical structure was equivalent to the product by a high temperature process or a monolithic wasteform consisting of a durable ceramic host matrix for immobilizing RE elements.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

    The isotope 99Tc is an important fission product generated from nuclear power production. Because of its long half-life (t1/2 = 2.13 ∙ 105 years) and beta-radiotoxicity (β⁻ = 292 keV), it is a major concern in the long-term management of spent nuclear fuel. In the spent nuclear fuel, Tc is present as an alloy with Mo, Ru, Rh, and Pd called the epsilon-phase, the relative amount of which increases with fuel burn-up. In some separation schemes for spent nuclear fuel, Tc would be separated from the spent fuel and disposed of in a durable waste form. Technetium waste forms under consideration include metallic alloys, oxide ceramics and borosilicate glass. In the development of a metallic waste form, after separation from the spent fuel, Tc would be converted to the metal, incorporated into an alloy and the resulting waste form stored in a repository. Metallic alloys under consideration include Tc–Zr alloys, Tc–stainless steel alloys and Tc–Inconel alloys (Inconel is an alloy of Ni, Cr and iron which is resistant to corrosion). To predict the long-term behavior of the metallic Tc waste form, understanding the corrosion properties of Tc metal and Tc alloys in various chemical environments is needed, but efforts to model the behavior of Tc metallic alloys are limited. One parameter that should also be considered in predicting the long-term behavior of the Tc waste form is the ingrowth of stable Ru that occurs from the radioactive decay of 99Tc (99Tc → 99Ru + β⁻). After a geological period of time, significant amounts of Ru will be present in the Tc and may affect its corrosion properties. Studying the effect of Ru on the corrosion behavior of Tc is also of importance. In this context, we studied the electrochemical behavior of Tc metal, Tc-Ni alloys (to model Tc-Inconel alloy) and Tc-Ru alloys in acidic media. The study of Tc-U alloys has also been performed in order to better understand the

  1. Crystallization behavior during melt-processing of ceramic waste forms

    Science.gov (United States)

    Tumurugoti, Priyatham; Sundaram, S. K.; Misture, Scott T.; Marra, James C.; Amoroso, Jake

    2016-05-01

    Multiphase ceramic waste forms based on natural mineral analogs are of great interest for their high chemical durability, radiation resistance, and thermodynamic stability. Melt-processed ceramic waste forms that leverage existing melter technologies will broaden the available disposal options for high-level nuclear waste. This work reports on the crystallization behavior in selected melt-processed ceramics for waste immobilization. The phase assemblage and evolution of hollandite, zirconolite, pyrochlore, and perovskite type structures during melt processing were studied using thermal analysis, x-ray diffraction, and electron microscopy. Samples prepared by melting followed by annealing and quenching were analyzed to determine and measure the progression of the phase assemblage. Samples were melted at 1500 °C and heat-treated at crystallization temperatures of 1285 °C and 1325 °C corresponding to exothermic events identified from differential scanning calorimetry measurements. Results indicate that the selected multiphase composition partially melts at 1500 °C with hollandite coexisting as crystalline phase. Perovskite and zirconolite phases crystallized from the residual melt at temperatures below 1350 °C. Depending on their respective thermal histories, different quenched samples were found to have different phase assemblages including phases such as perovskite, zirconolite and TiO2.

  2. Preliminary waste form characteristics report Version 1.0. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.; Leider, H.R. [eds.

    1991-10-11

    This report focuses on radioactive waste form characteristics that will be used to design a waste package and an engineered barrier system (EBS) for a suitable repository as part of the Yucca Mountain Project. The term waste form refers to irradiated reactor fuel, other high-level waste (HLW) in various physical forms, and other radioactive materials (other than HLW) which are received for emplacement in a geologic repository. Any encapsulating of stabilizing matrix is also referred to as a waste form.

  3. Low sintering temperature glass waste forms for sequestering radioactive iodine

    Science.gov (United States)

    Nenoff, Tina M.; Krumhansl, James L.; Garino, Terry J.; Ockwig, Nathan W.

    2012-09-11

    Materials and methods of making low-sintering-temperature glass waste forms that sequester radioactive iodine in a strong and durable structure. First, the iodine is captured by an adsorbant, which forms an iodine-loaded material, e.g., AgI, AgI-zeolite, AgI-mordenite, Ag-silica aerogel, ZnI.sub.2, CuI, or Bi.sub.5O.sub.7I. Next, particles of the iodine-loaded material are mixed with powdered frits of low-sintering-temperature glasses (comprising various oxides of Si, B, Bi, Pb, and Zn), and then sintered at a relatively low temperature, ranging from 425.degree. C. to 550.degree. C. The sintering converts the mixed powders into a solid block of a glassy waste form, having low iodine leaching rates. The vitrified glassy waste form can contain as much as 60 wt % AgI. A preferred glass, having a sintering temperature of 500.degree. C. (below the silver iodide sublimation temperature of 500.degree. C.) was identified that contains oxides of boron, bismuth, and zinc, while containing essentially no lead or silicon.

  4. Estimation of centerline temperature of the waste form for the rare earth waste generated from pyrochemical process

    Science.gov (United States)

    Choi, Jung-Hoon; Eun, Hee-Chul; Lee, Tae-Kyo; Lee, Ki-Rak; Han, Seung-Youb; Jeon, Min-Ku; Park, Hwan-Seo; Ahn, Do-Hee

    2017-01-01

    Estimation of centerline temperature of nuclear glass waste form for each waste stream is very essential in the period of storage because the centerline temperature being over its glass transition temperature results in the increase of leaching rate of radioactive nuclides due to the devitrification of glass waste form. Here, to verify the effects of waste form diameter and transuranic element content in the rare earth waste on the centerline temperature of the waste form, the surrogate rare earth glass waste generated from pyrochemical process was immobilized with SiO2sbnd Al2O3sbnd B2O3 glass frit system, and thermal properties of the rare earth glass waste form were determined by thermomechanical analysis and thermal conductivity analysis. The estimation of centerline temperature was carried out using the experimental thermal data and steady-state conduction equation in a long and solid cylinder type waste form. It was revealed that thermal stability of waste form in case of 0.3 m diameter was not affected by the TRU content even in the case of 80% TRU recovery ratio in the electrowinning process, meaning that the waste form of 0.3 m diameter is thermally stable due to the low centerline temperature relative to its glass transition temperature of the rare earth glass waste form.

  5. Commercial high-level-waste management: options and economics. A comparative analysis of the ceramic and glass waste forms

    Energy Technology Data Exchange (ETDEWEB)

    McKisson, R.L.; Grantham, L.F.; Guon, J.; Recht, H.L.

    1983-02-25

    Results of an estimate of the waste management costs of the commercial high-level waste from a 3000 metric ton per year reprocessing plant show that the judicious use of the ceramic waste form can save about $2 billion during a 20-year operating campaign relative to the use of the glass waste form. This assumes PWR fuel is processed and the waste is encapsulated in 0.305-m-diam canisters with ultimate emplacement in a BWIP-type horizontal-borehole repository. The estimated total cost (capital and operating) of the management in the ceramic form is $2.0 billion, and that of the glass form is $4.0 billion. Waste loading and waste form density are the driving factors in that the low-waste loading (25%) and relatively low density (3.1 g/cm/sup 3/) characteristic of the glass form require several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 g/cm/sup 3/) characteristic of the glass form requires several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 g/cm/sup 3/. The minimum-cost ceramic waste form has a 60 wt. % waste loading of commercial high-level waste and requires 25 years storage before emplacement in basalt with delayed backfill. Because of the process flexibility allowed by the availability of the high-waste loading of the ceramic form, the intermediate-level liquid waste stream can be mixed with the high-level liquid waste stream and economically processed and emplaced. The cost is greater by $0.3 billion than that of the best high-level liquid waste handling process sequence ($2.3 billion vs $2.0 billion), but this difference is less than the cost of the separate disposal of the intermediate-level liquid waste.

  6. Cesium incorporation in hollandite-rich multiphasic ceramic waste forms

    Science.gov (United States)

    Tumurugoti, P.; Clark, B. M.; Edwards, D. J.; Amoroso, Jake; Sundaram, S. K.

    2017-02-01

    Hollandite-rich multiphase waste form compositions processed by melt-solidification and spark plasma sintering (SPS) were characterized, compared, and validated for nuclear waste incorporation. Phase identification by x-ray diffraction (XRD) and electron back-scattered diffraction (EBSD) confirmed hollandite as the major phase present in these samples along with perovskite, pyrochlore and zirconolite. Distribution of selected elements observed by wavelength dispersive spectroscopy (WDS) maps indicated that Cs formed a secondary phase during SPS processing, which was considered undesirable. On the other hand, Cs partitioned into the hollandite phase in melt-processed samples. Further analysis of hollandite structure in melt-processed composition by selected area electron diffraction (SAED) revealed ordered arrangement of tunnel ions (Ba/Cs) and vacancies, suggesting efficient Cs incorporation into the lattice.

  7. Material Recover and Waste Form Development--2016 Accomplishments

    Energy Technology Data Exchange (ETDEWEB)

    Todd, Terry A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vienna, John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Paviet, Patricia [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    The Material Recovery and Waste Form Development (MRWFD) Campaign under the U.S. Department of Energy (DOE) Fuel Cycle Technologies (FCT) Program is responsible for developing advanced separation and waste form technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress (April 2010). This MRWFD accomplishments report summarizes the results of the research and development (R&D) efforts performed within MRWFD in Fiscal Year (FY) 2016. Each section of the report contains an overview of the activities, results, technical point of contact, applicable references, and documents produced during the FY. This report briefly outlines campaign management and integration activities but primarily focuses on the many technical accomplishments of FY 2016. The campaign continued to use an engineering-driven, science-based approach to maintain relevance and focus.

  8. Technical viability and development needs for waste forms and facilities

    Energy Technology Data Exchange (ETDEWEB)

    Pegg, I.; Gould, T.

    1996-05-01

    The objective of this breakout session was to provide a forum to discuss technical issues relating to plutonium-bearing waste forms and their disposal facilities. Specific topics for discussion included the technical viability and development needs associated with the waste forms and/or disposal facilities. The expected end result of the session was an in-depth (so far as the limited time would allow) discussion of key issues by the session participants. The session chairs expressed allowance for, and encouragement of, alternative points of view, as well as encouragement for discussion of any relevant topics not addressed in the paper presentations. It was not the intent of this session to recommend or advocate any one technology over another.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-08-01

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

  10. The Ceramic Waste Form Process at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Ken Bateman; Stephen Priebe

    2006-08-01

    The treatment of spent nuclear fuel for disposition using an electrometallurgical technique results in two high-level waste forms: a ceramic waste form (CWF) and a metal waste form (MWF). The CWF is a composite of sodalite and glass, which stabilizes the active fission products (alkali, alkaline earths, and rare earths) and transuranic (TRU) elements. Reactive metal fuel constituents, including all the TRU metals and the majority of the fission products remain in the salt as chlorides and are processed into the CWF. The solidified salt is containerized and transferred to the CWF process where it is ground in an argon atmosphere. Zeolite 4A is dried in a mechanically-fluidized dryer to about 0.1 wt% moisture and ground to a particle-size range of 45µ to 250µ. The salt and zeolite are mixed in a V-mixer and heated to 500°C for about 18 hours. During this process, the salt occludes into the structure of the zeolite. The salt-loaded zeolite (SLZ) is cooled and then mixed with borosilicate glass frit with a comparable particle-size range. The SLZ/glass mixture is transferred to a crucible, which is placed in a furnace and heated to 925°C. During this process, known as pressureless consolidation, the zeolite is converted to the final sodalite form and the glass thoroughly encapsulates the sodalite, producing a dense, leach-resistant final waste form. During the last several years, changes have occurred to the process, including: particle size of input materials and conversion from hot isostatic pressing to pressureless consolidation, This paper is intended to provide the current status of the CWF process focusing on the adaptation to pressureless consolidation. Discussions will include impacts of particle size on final waste form and the pressureless consolidation cycle. A model will be presented that shows the heating and cooling cycles and the effect of radioactive decay heat on the amount of fission products that can be incorporated into the CWF.

  11. 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.

  12. Description of DWPF reference waste form and canister

    Energy Technology Data Exchange (ETDEWEB)

    1981-06-01

    This document describes the reference waste form and canister for the Defense Waste Processing Facility (DWPF). The facility is planned for location at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1983. The reference canister is fabricated of 24-in.-OD 304L stainless steel pipe with a dished bottom, domed head, and lifting and welding flanges on the head neck. The overall canister length is 9 ft 10 in., with a wall thickness of 3/8-in. (schedule 20 pipe). The canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected to ensure that a filled canister with its shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be generally compatible with preliminary assessments of repository requirements. The reference waste form is borosilicate glass containing approximately 28 wt % sludge oxides with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains approximately 58% SiO/sub 2/ and 15% B/sub 2/O/sub 3/. This composition results in a low average leachability in the waste form of approximately 5 x 10/sup -9/ g/cm/sup 2/-day based on /sup 137/Cs over 365 days in 25/sup 0/C water. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approximately 425 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the Stage 1 and Stage 2 processes. The radionuclide content of the canister is about 150,000 curies, with a radiation level of 2 x 10/sup 4/ rem/hour at 1 cm.

  13. Pyrochlore as nuclear waste form. Actinide uptake and chemical stability

    Energy Technology Data Exchange (ETDEWEB)

    Finkeldei, Sarah Charlotte

    2015-07-01

    Radioactive waste is generated by many different technical and scientific applications. For the past decades, different waste disposal strategies have been considered. Several questions on the waste disposal strategy remain unanswered, particularly regarding the long-term radiotoxicity of minor actinides (Am, Cm, Np), plutonium and uranium. These radionuclides mainly arise from high level nuclear waste (HLW), specific waste streams or dismantled nuclear weapons. Although many countries have opted for the direct disposal of spent fuel, from a scientific and technical point of view it is imperative to pursue alternative waste management strategies. Apart from the vitrification, especially for trivalent actinides and Pu, crystalline ceramic waste forms are considered. In contrast to glasses, crystalline waste forms, which are chemically and physically highly stable, allow the retention of radionuclides on well-defined lattice positions within the crystal structure. Besides polyphase ceramics such as SYNROC, single phase ceramics are considered as tailor made host phases to embed a specific radionuclide or a specific group. Among oxidic single phase ceramics pyrochlores are known to have a high potential for this application. This work examines ZrO{sub 2} based pyrochlores as potential nuclear waste forms, which are known to show a high aqueous stability and a high tolerance towards radiation damage. This work contributes to (1) understand the phase stability field of pyrochlore and consequences of non-stoichiometry which leads to pyrochlores with mixed cationic sites. Mixed cationic occupancies are likely to occur in actinide-bearing pyrochlores. (2) The structural uptake of radionuclides themselves was studied. (3) The chemical stability and the effect of phase transition from pyrochlore to defect fluorite were probed. This phase transition is important, as it is the result of radiation damage in ZrO{sub 2} based pyrochlores. ZrO{sub 2} - Nd{sub 2}O{sub 3} pellets

  14. Dilute condition corrosion behavior of glass-ceramic waste form

    Science.gov (United States)

    Crum, Jarrod V.; Neeway, James J.; Riley, Brian J.; Zhu, Zihua; Olszta, Matthew J.; Tang, Ming

    2016-12-01

    Borosilicate glass-ceramics are being developed to immobilize high-level waste generated by aqueous reprocessing into a stable waste form. The corrosion behavior of this multiphase waste form is expected to be complicated by multiple phases and crystal-glass interfaces. A modified single-pass flow-through test was performed on polished monolithic coupons at a neutral pH (25 °C) and 90 °C for 33 d. The measured glass corrosion rates by micro analysis in the samples ranged from 0.019 to 0.29 g m-2 d-1 at a flow rate per surface area = 1.73 × 10-6 m s-1. The crystal phases (oxyapatite and Ca-rich powellite) corroded below quantifiable rates, by micro analysis. While, Ba-rich powellite corroded considerably in O10 sample. The corrosion rates of C1 and its replicate C20 were elevated an order of magnitude by mechanical stresses at crystal-glass interface caused by thermal expansion mismatch during cooling and unique morphology (oxyapatite clustering).

  15. Radiation damage studies related to nuclear waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Gray, W.J.; Wald, J.W.; Turcotte, R.P.

    1981-12-01

    Much of the previously reported work on alpha radiation effects on crystalline phases of importance to nuclear waste forms has been derived from radiation effects studies of composite waste forms. In the present work, two single-phase crystalline materials, Gd/sub 2/Ti/sub 2/O/sub 7/ (pyrochlore) and CaZrTi/sub 2/O/sub 7/ (zirconolite), of relative importance to current waste forms were studied independently by doping with /sup 244/Cm at the 3 wt % level. Changes in the crystalline structure measured by x-ray diffraction as a function of dose show that damage ingrowth follows an expected exponential relationship of the form ..delta..V/V/sub 0/ = A(1-exp(-BD)). In both cases, the materials became x-ray amorphous before the estimated saturation value was reached. The predicted magnitudes of the unit cell volume changes at saturation are 5.4% and 3.5%, respectively, for Gd/sub 2/Ti/sub 2/O/sub 7/ and CaZrTi/sub 2/O/sub 7/. The later material exhibited anisotropic behavior in which the expansion of the monoclinic cell in the c/sub 0/ direction was over five times that of the a/sub 0/ direction. The effects of transmutations on the properties of high-level waste solids have not been studied until now because of the long half-lives of the important fission products. This problem was circumvented in the present study by preparing materials containing natural cesium and then irradiating them with neutrons to produce /sup 134/Cs, which has only a 2y half-life. The properties monitored at about one year intervals following irradiation have been density, leach rate and microstructure. A small amount of x-ray diffraction work has also been done. Small changes in density and leach rate have been observed for some of the materials, but they were not large enough to be of any consequence for the final disposal of high level wastes.

  16. Crystalline ceramics: Waste forms for the disposal of weapons plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, R.C.; Lutze, W. [New Mexico Univ., Albuquerque, NM (United States); Weber, W.J. [Pacific Northwest Lab., Richland, WA (United States)

    1995-05-01

    At present, there are three seriously considered options for the disposition of excess weapons plutonium: (i) incorporation, partial burn-up and direct disposal of MOX-fuel; (ii) vitrification with defense waste and disposal as glass ``logs``; (iii) deep borehole disposal (National Academy of Sciences Report, 1994). The first two options provide a safeguard due to the high activity of fission products in the irradiated fuel and the defense waste. The latter option has only been examined in a preliminary manner, and the exact form of the plutonium has not been identified. In this paper, we review the potential for the immobilization of plutonium in highly durable crystalline ceramics apatite, pyrochlore, monazite and zircon. Based on available data, we propose zircon as the preferred crystalline ceramic for the permanent disposition of excess weapons plutonium.

  17. Colloid formation during waste form reaction: implications for nuclear waste disposal

    Science.gov (United States)

    Bates, J. K.; Bradley, J.; Teetsov, A.; Bradley, C. R.; ten Brink, Marilyn Buchholtz

    1992-01-01

    Insoluble plutonium- and americium-bearing colloidal particles formed during simulated weathering of a high-level nuclear waste glass. Nearly 100 percent of the total plutonium and americium in test ground water was concentrated in these submicrometer particles. These results indicate that models of actinide mobility and repository integrity, which assume complete solubility of actinides in ground water, underestimate the potential for radionuclide release into the environment. A colloid-trapping mechanism may be necessary for a waste repository to meet long-term performance specifications.

  18. Fundamental Science-Based Simulation of Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Gao, Fei; Sun, Xin; Khaleel, Mohammad A.

    2010-10-04

    This report presents a hierarchical multiscale modeling scheme based on two-way information exchange. To account for all essential phenomena in waste forms over geological time scales, the models have to span length scales from nanometer to kilometer and time scales from picoseconds to millenia. A single model cannot cover this wide range and a multi-scale approach that integrates a number of different at-scale models is called for. The approach outlined here involves integration of quantum mechanical calculations, classical molecular dynamics simulations, kinetic Monte Carlo and phase field methods at the mesoscale, and continuum models. The ultimate aim is to provide science-based input in the form of constitutive equations to integrated codes. The atomistic component of this scheme is demonstrated in the promising waste form xenotime. Density functional theory calculations have yielded valuable information about defect formation energies. This data can be used to develop interatomic potentials for molecular dynamics simulations of radiation damage. Potentials developed in the present work show a good match for the equilibrium lattice constants, elastic constants and thermal expansion of xenotime. In novel waste forms, such as xenotime, a considerable amount of data needed to validate the models is not available. Integration of multiscale modeling with experimental work is essential to generate missing data needed to validate the modeling scheme and the individual models. Density functional theory can also be used to fill knowledge gaps. Key challenges lie in the areas of uncertainty quantification, verification and validation, which must be performed at each level of the multiscale model and across scales. The approach used to exchange information between different levels must also be rigorously validated. The outlook for multiscale modeling of wasteforms is quite promising.

  19. DuraLith geopolymer waste form for Hanford secondary waste: Correlating setting behavior to hydration heat evolution

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Hui; Gong, Weiliang, E-mail: gongw@vsl.cua.edu; Syltebo, Larry; Lutze, Werner; Pegg, Ian L.

    2014-08-15

    Highlights: • Quantitative correlations firstly established for cementitious waste forms. • Quantitative correlations firstly established for geopolymeric materials. • Ternary DuraLith geopolymer waste forms for Hanford radioactive wastes. • Extended setting times which improve workability for geopolymer waste forms. • Reduced hydration heat release from DuraLith geopolymer waste forms. - Abstract: The binary furnace slag-metakaolin DuraLith geopolymer waste form, which has been considered as one of the candidate waste forms for immobilization of certain Hanford secondary wastes (HSW) from the vitrification of nuclear wastes at the Hanford Site, Washington, was extended to a ternary fly ash-furnace slag-metakaolin system to improve workability, reduce hydration heat, and evaluate high HSW waste loading. A concentrated HSW simulant, consisting of more than 20 chemicals with a sodium concentration of 5 mol/L, was employed to prepare the alkaline activating solution. Fly ash was incorporated at up to 60 wt% into the binder materials, whereas metakaolin was kept constant at 26 wt%. The fresh waste form pastes were subjected to isothermal calorimetry and setting time measurement, and the cured samples were further characterized by compressive strength and TCLP leach tests. This study has firstly established quantitative linear relationships between both initial and final setting times and hydration heat, which were never discovered in scientific literature for any cementitious waste form or geopolymeric material. The successful establishment of the correlations between setting times and hydration heat may make it possible to efficiently design and optimize cementitious waste forms and industrial wastes based geopolymers using limited testing results.

  20. Engineering-Scale Demonstration of DuraLith and Ceramicrete Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Josephson, Gary B.; Westsik, Joseph H.; Pires, Richard P.; Bickford, Jody; Foote, Martin W.

    2011-09-23

    To support the selection of a waste form for the liquid secondary wastes from the Hanford Waste Immobilization and Treatment Plant, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing on four candidate waste forms. Two of the candidate waste forms have not been developed to scale as the more mature waste forms. This work describes engineering-scale demonstrations conducted on Ceramicrete and DuraLith candidate waste forms. Both candidate waste forms were successfully demonstrated at an engineering scale. A preliminary conceptual design could be prepared for full-scale production of the candidate waste forms. However, both waste forms are still too immature to support a detailed design. Formulations for each candidate waste form need to be developed so that the material has a longer working time after mixing the liquid and solid constituents together. Formulations optimized based on previous lab studies did not have sufficient working time to support large-scale testing. The engineering-scale testing was successfully completed using modified formulations. Further lab development and parametric studies are needed to optimize formulations with adequate working time and assess the effects of changes in raw materials and process parameters on the final product performance. Studies on effects of mixing intensity on the initial set time of the waste forms are also needed.

  1. 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.

  2. Naturally occurring crystalline phases: analogues for radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Haaker, R.F.; Ewing, R.C.

    1981-01-01

    Naturally occurring mineral analogues to crystalline phases that are constituents of crystalline radioactive waste forms provide a basis for comparison by which the long-term stability of these phases may be estimated. The crystal structures and the crystal chemistry of the following natural analogues are presented: baddeleyite, hematite, nepheline; pollucite, scheelite;sodalite, spinel, apatite, monazite, uraninite, hollandite-priderite, perovskite, and zirconolite. For each phase in geochemistry, occurrence, alteration and radiation effects are described. A selected bibliography for each phase is included.

  3. Chromium speciation in hazardous, cement-based waste forms

    Science.gov (United States)

    Lee, J. F.; Bajt, S.; Clark, S. B.; Lamble, G. M.; Langton, C. A.; Oji, L.

    1995-02-01

    XANES and EXAFS techniques were used to determine the oxidation states and local structural environment of Cr in cement-based waste forms. Results show that Cr in untreated Portland cement formulations remains as toxic Cr 6+, while slag additives to the cement reduce Cr 6+ to the less toxic, less mobile Cr 3+ species. EXAFS analysis suggests that the Cr 6+ species is surrounded by four nearest oxygen atoms, while the reduced Cr 3+ sp ecies is surrounded by six oxygen atoms. The fitted CrO bond lengths for Cr 6+ and Cr 3+ species are around 1.66 and 1.98 Å, respectively.

  4. Garnet nuclear waste forms – Solubility at repository conditions

    Energy Technology Data Exchange (ETDEWEB)

    Caporuscio, F.A., E-mail: floriec@lanl.gov [EES-14, Los Alamos National Laboratory, NM 87545 (United States); Scott, B.L. [MPA-MSID, Los Alamos National Laboratory, NM 87545 (United States); Xu, H. [EES-14, Los Alamos National Laboratory, NM 87545 (United States); Feller, R.K. [Effect Materials Research Group, BASF Corporation, 500 White Plains Road, Tarrytown, NY 10591 (United States)

    2014-01-15

    Highlights: • Rare-earth elements are a significant waste stream produced by nuclear fuel cycles. • Suitability of garnets as potential waste forms. • Single-crystal X-ray structural refinements for grossular, LuAG and YAG. • Garnets have low solubility, flexible crystal structure to take on large cations. • Demonstrate garnets are potentially robust waste forms for radioactive REE. -- Abstract: Radioactive rare-earth elements (REEs) constitute a significant waste stream produced from modified open and full nuclear fuel cycles. Immobilization of these REE radionuclides is thus important for sustainable nuclear energy growth. In this work, we investigated the suitability of garnets as potential waste forms for REEs by measuring their aqueous stability at repository conditions. Three garnet samples, including one natural grossular (Ca{sub 3}Al{sub 2}Si{sub 3}O{sub 12}) and two synthetic phases (LuAG – Lu{sub 3}Al{sub 5}O{sub 12} and YAG – Y{sub 3}Al{sub 5}O{sub 12}), were studied. Single-crystal X-ray structural refinements show that the unit-cell volumes increase from 1657.19 Å{sup 3} for grossular to 1679.8 Å{sup 3} for LuAG and to 1721.7 Å{sup 3} for YAG. This trend is due to increases in ionic radii in both the 8-coordinated X (from Ca to Lu to Y) and 4-coordinated Z (from Si to Al) cations. Hydrothermal experiments of the three samples were performed at 200 °C and 150 bar for 4 weeks using water and brine solutions to evaluate their solubility. The natural grossular sample exhibited Al leach rates ranging from 2.5 × 10{sup −4} to 6.43 × 10{sup −5} g/L·day and Ca leach rates from 1.39 × 10{sup −3} to 4.57 × 10{sup −3} g/L·day, indicating incongruent nature of the cation dissolution. The LuAG sample exhibited Lu leach rates of 3.73 × 10{sup −4} to 2.19 × 10{sup −4} g/L·day, and the YAG sample had Y leach rates of 1.29 × 10{sup −4} to 5.64 × 10{sup −5} g/L·day. Although these samples are generally more soluble in

  5. Radiation damage of hollandite in multiphase ceramic waste forms

    Science.gov (United States)

    Clark, Braeden M.; Tumurgoti, Priyatham; Sundaram, S. K.; Amoroso, Jake W.; Marra, James C.; Shutthanandan, Vaithiyalingam; Tang, Ming

    2017-10-01

    Radiation damage was simulated in multiphase titanate-based ceramic waste forms using an ion accelerator to generate high energy alpha particles (He+) and an ion implanter to generate 7 MeV gold (Au3+) particles. X-ray diffraction and transmission electron microscopy were used to characterize the damaged surfaces and nearby regions. Simulated multiphase ceramic waste forms were prepared using two processing methods: spark plasma sintering and melt-processing. Both processing methods produced ceramics with similar phase assemblages consisting of hollandite-, zirconolite/pyrochlore-, and perovskite-type phases. The measured heavy ion (Au3+) penetration depth was less in spark plasma sintered samples than in melt-processed samples. Structural breakdown of the hollandite phase occurred under He+ irradiation indicated by the presence of x-ray diffraction peaks belonging to TiO2, BaTiO5, and other hollandite related phases (Ba2Ti9O20). The composition of the constituent hollandite phase affected the extent of damage induced by Au3+ ions.

  6. Material Recovery and Waste Form Development FY 2015 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Todd, Terry Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Braase, Lori Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-11-01

    The Material Recovery and Waste Form Development (MRWFD) Campaign under the U.S. Department of Energy (DOE) Fuel Cycle Technologies (FCT) Program is responsible for developing advanced separation and waste form technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The FY 2015 Accomplishments Report provides a highlight of the results of the research and development (R&D) efforts performed within the MRWFD Campaign in FY-14. Each section contains a high-level overview of the activities, results, technical point of contact, applicable references, and documents produced during the fiscal year. This report briefly outlines campaign management and integration activities, but primarily focuses on the many technical accomplishments made during FY-15. The campaign continued to utilize an engineering driven-science-based approach to maintain relevance and focus. There was increased emphasis on development of technologies that support near-term applications that are relevant to the current once-through fuel cycle.

  7. Candidate waste forms for immobilisation of waste chloride salt from pyroprocessing of spent nuclear fuel

    Science.gov (United States)

    Vance, E. R.; Davis, J.; Olufson, K.; Chironi, I.; Karatchevtseva, I.; Farnan, I.

    2012-01-01

    Sodalite/glass bodies prepared by hot isostatic pressing (HIPing) at ˜850 °C/100 MPa are candidates for immobilising fission product-bearing waste KCl-LiCl pyroprocessing salts. To study the capacity of sodalite to structurally incorporate such pyroprocessing salts, K, Li, Cs, Sr, Ba and La were individually targeted for substitution in a Na site in sodalite (Na vacancies targeted as charge compensators for alkaline and rare earths) and studied by X-ray diffraction and scanning electron microscopy after sintering in the range of 800-1000 °C. K and Li appeared to enter the sodalite, but Cs, Sr and Ba formed aluminosilicate phases and La formed an oxyapatite phase. However these non-sodalite phases have reasonable resistance to water leaching. Pure chlorapatite gives superior leach resistance to sodalite, and alkalis, alkaline and rare earth ions are generally known to enter chlorapatite, but attempts to incorporate simulated waste salt formulations into HIPed chlorapatite-based preparations or to substitute Cs alone into the structure of Ca-based chlorapatite were not successful on the basis of scanning electron microscopy. The materials exhibited severe water leachability, mainly in regard to Cs release. Attempts to substitute Cs into Ba- and Sr-based chlorapatites also did not look encouraging. Consequently the use of apatite alone to retain fission product-bearing waste pyroprocessing salts from electrolytic nuclear fuel reprocessing is problematical, but chlorapatite glass-ceramics may be feasible, albeit with reduced waste loadings. Spodiosite, Ca 2(PO 4)Cl, does not appear to be suitable for incorporation of Cl-bearing waste containing fission products.

  8. The durability of single, dual, and multiphase titanate ceramic waste forms for nuclear waste immobilization

    Science.gov (United States)

    Harkins, Devin J. H.

    A significant amount of the energy used in the United States comes from nuclear power, which produces a large amount of waste materials. Recycling nuclear waste is possible, but requires a way to permanently fix the unusable radionuclides remaining from the recycling process in a stable, leach resistant structure. Multiphase titanate ceramic waste forms are one promising option under consideration. However, there is insufficient work on the long term corrosion of the individual phases, as well as the multiphase systems of these ceramics. These multiphase titanate ceramic waste forms have three targeted phases: hollandite, pyrochlore, and zirconolite. Hollandite is a promising candidate for the incorporation of Cs, while pyrochlore is readily formed with lanthanides, such as Nd, the most prevalent lanthanide in the waste stream. The third targeted phase, zirconolite, is for the incorporation of zirconium and the actinides. This work looks into the formation of single phase systems of lanthanide titanates, formation of dual phase systems of Ga doped Ba hollandites and Nd titanate, durability of single phase hollandites and multiphase model systems using Vapor Hydration Testing (ASTM C 1663-09), dissolution of dual phase systems of Ga doped Ba hollandites and Nd titanate using Product Consistency Testing (ASTM C 1285-02), as well investigating how grain size affects amount of alterative phases formed using Vapor Hydration Testing. The dual phase systems of hollandites and Nd titanate show significant amounts of secondary phases forming, heavily influenced by the composition of hollandite used in the systems. The most significant phase present was BaNd2Ti5O14. This phase proves to be problematic due to the degradation to the hollandite structure. Using Vapor Hydration Testing to investigate single and multiphase systems presented many some possible alteration phases that could occur in the long term aging of these ceramics. Most notably, Cs rich phases were found in

  9. Impeding 99Tc(IV) mobility in novel waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Mal Soon; Um, Wooyong; Wang, Guohui; Kruger, Albert A.; Lukens, Wayne W.; Rousseau, Roger J.; Glezakou, Vassiliki Alexandra

    2016-06-30

    Technetium (99Tc) is a long-lived radioactive fission product whose mobility in the subsurface is largely governed by its oxidation state1. Immobilization of Tc in mineral substrates is crucial for radioactive waste management and environmental remediation. Tc(IV) incorporation in spinels2, 3 has been proposed as a novel method to increase Tc retention in glass waste forms. However, experiments with Tc-magnetite under high temperature and oxic conditions showed re-oxidation of Tc(IV) to volatile pertechnetate Tc(VII)O4-.4, 5 Here we address this problem with large-scale ab initio molecular dynamics simulations and propose that elevated temperatures, 1st row transition metal dopants can significantly enhance Tc retention in the order Co > Zn > Ni. Experiments with doped spinels at T=700 ºC provided quantitative confirmation of increased Tc retention in the same order predicted by theory. This work highlights the power of modern state-of-the-art simulations to provide essential insights and generate bottom-up design criteria of complex oxide materials at elevated temperatures.

  10. Impeding 99Tc(IV) mobility in novel waste forms

    Science.gov (United States)

    Lee, Mal-Soon; Um, Wooyong; Wang, Guohui; Kruger, Albert A.; Lukens, Wayne W.; Rousseau, Roger; Glezakou, Vassiliki-Alexandra

    2016-06-01

    Technetium (99Tc) is an abundant, long-lived radioactive fission product whose mobility in the subsurface is largely governed by its oxidation state. Tc immobilization is crucial for radioactive waste management and environmental remediation. Tc(IV) incorporation in spinels has been proposed as a novel method to increase Tc retention in glass waste forms during vitrification. However, experiments under high-temperature and oxic conditions show reoxidation of Tc(IV) to volatile pertechnetate, Tc(VII). Here we examine this problem with ab initio molecular dynamics simulations and propose that, at elevated temperatures, doping with first row transition metal can significantly enhance Tc retention in magnetite in the order Co>Zn>Ni. Experiments with doped spinels at 700 °C provide quantitative confirmation of the theoretical predictions in the same order. This work highlights the power of modern, state-of-the-art simulations to provide essential insights and generate theory-inspired design criteria of complex materials at elevated temperatures.

  11. Alternative Electrochemical Salt Waste Forms, Summary of FY2010 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Rieck, Bennett T.; Crum, Jarrod V.; Matyas, Josef; McCloy, John S.; Sundaram, S. K.; Vienna, John D.

    2010-08-01

    In FY2009, PNNL performed scoping studies to qualify two waste form candidates, tellurite (TeO2-based) glasses and halide minerals, for the electrochemical waste stream for further investigation. Both candidates showed promise with acceptable PCT release rates and effective incorporation of the 10% fission product waste stream. Both candidates received reprisal for FY2010 and were further investigated. At the beginning of FY2010, an in-depth literature review kicked off the tellurite glasses study. The review was aimed at ascertaining the state-of-the-art for chemical durability testing and mixed chloride incorporation for tellurite glasses. The literature review led the authors to 4 unique binary and 1 unique ternary systems for further investigation which include TeO2 plus the following: PbO, Al2O3-B2O3, WO3, P2O5, and ZnO. Each system was studied with and without a mixed chloride simulated electrochemical waste stream and the literature review provided the starting points for the baseline compositions as well as starting points for melting temperature, compatible crucible types, etc. The most promising glasses in each system were scaled up in production and were analyzed with the Product Consistency Test, a chemical durability test. Baseline and PCT glasses were analyzed to determine their state, i.e., amorphous, crystalline, phase separated, had undissolved material within the bulk, etc. Conclusions were made as well as the proposed direction for FY2011 plans. Sodalite was successfully synthesized by the sol-gel method. The vast majority of the dried sol-gel consisted of sodalite with small amounts of alumino-silicates and unreacted salt. Upon firing the powders made by sol-gel, the primary phase observed was sodalite with the addition of varying amounts of nepheline, carnegieite, lithium silicate, and lanthanide oxide. The amount of sodalite, nepheline, and carnegieite as well as the bulk density of the fired pellets varied with firing temperature, sol

  12. MODELING SOLIDIFICATION-INDUCED STRESSES IN CERAMIC WASTE FORMS CONTAINING NUCLEAR WASTES

    Energy Technology Data Exchange (ETDEWEB)

    Charles W. Solbrig; Kenneth J. Bateman

    2010-11-01

    The goal of this work is to produce a ceramic waste form (CWF) that permanently occludes radioactive waste. This is accomplished by absorbing radioactive salts into zeolite, mixing with glass frit, heating to a molten state 915 C to form a sodalite glass matrix, and solidifying for long-term storage. Less long term leaching is expected if the solidifying cooling rate doesn’t cause cracking. In addition to thermal stress, this paper proposes that a stress is formed during solidification which is very large for fast cooling rates during solidification and can cause severe cracking. A solidifying glass or ceramic cylinder forms a dome on the cylinder top end. The temperature distribution at the time of solidification causes the stress and the dome. The dome height, “the length deficit,” produces an axial stress when the solid returns to room temperature with the inherent outer region in compression, the inner in tension. Large tensions will cause cracking of the specimen. The temperature deficit, derived by dividing the length deficit by the coefficient of thermal expansion, allows solidification stress theory to be extended to the circumferential stress. This paper derives the solidification stress theory, gives examples, explains how to induce beneficial stresses, and compares theory to experimental data.

  13. DuraLith geopolymer waste form for Hanford secondary waste: correlating setting behavior to hydration heat evolution.

    Science.gov (United States)

    Xu, Hui; Gong, Weiliang; Syltebo, Larry; Lutze, Werner; Pegg, Ian L

    2014-08-15

    The binary furnace slag-metakaolin DuraLith geopolymer waste form, which has been considered as one of the candidate waste forms for immobilization of certain Hanford secondary wastes (HSW) from the vitrification of nuclear wastes at the Hanford Site, Washington, was extended to a ternary fly ash-furnace slag-metakaolin system to improve workability, reduce hydration heat, and evaluate high HSW waste loading. A concentrated HSW simulant, consisting of more than 20 chemicals with a sodium concentration of 5 mol/L, was employed to prepare the alkaline activating solution. Fly ash was incorporated at up to 60 wt% into the binder materials, whereas metakaolin was kept constant at 26 wt%. The fresh waste form pastes were subjected to isothermal calorimetry and setting time measurement, and the cured samples were further characterized by compressive strength and TCLP leach tests. This study has firstly established quantitative linear relationships between both initial and final setting times and hydration heat, which were never discovered in scientific literature for any cementitious waste form or geopolymeric material. The successful establishment of the correlations between setting times and hydration heat may make it possible to efficiently design and optimize cementitious waste forms and industrial wastes based geopolymers using limited testing results.

  14. Radionuclide Incorporation and Long Term Performance of Apatite Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jianwei [Louisiana State Univ., Baton Rouge, LA (United States); Lian, Jie [Rensselaer Polytechnic Inst., Troy, NY (United States); Gao, Fei [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-01-04

    This project aims to combines state-of-the-art experimental and characterization techniques with atomistic simulations based on density functional theory (DFT) and molecular dynamics (MD) simulations. With an initial focus on long-lived I-129 and other radionuclides such as Cs, Sr in apatite structure, specific research objectives include the atomic scale understanding of: (1) incorporation behavior of the radionuclides and their effects on the crystal chemistry and phase stability; (2) stability and microstructure evolution of designed waste forms under coupled temperature and radiation environments; (3) incorporation and migration energetics of radionuclides and release behaviors as probed by DFT and molecular dynamics (MD) simulations; and (4) chemical durability as measured in dissolution experiments for long term performance evaluation and model validation.

  15. Progress in forming bottom barriers under waste sites

    Energy Technology Data Exchange (ETDEWEB)

    Carter, E.E. [Carter Technologies, Sugar Land, TX (United States)

    1997-12-31

    The paper describes an new method for the construction, verification, and maintenance of underground vaults to isolate and contain radioactive burial sites without excavation or drilling in contaminated areas. The paper begins with a discussion of previous full-scale field tests of horizontal barrier tools which utilized high pressure jetting technology. This is followed by a discussion of the TECT process, which cuts with an abrasive cable instead of high pressure jets. The new method is potentially applicable to more soil types than previous methods and can form very thick barriers. Both processes are performed from the perimeter of a site and require no penetration or disturbance of the active waste area. The paper also describes long-term verification methods to monitor barrier integrity passively.

  16. Waste form characteristics report, revision 1.3

    Energy Technology Data Exchange (ETDEWEB)

    Leider, H.R.; Stout, R.B.

    1998-07-01

    This Waste Form Characteristics Report (WFCR) update, Version 1.3, incorporates substantial additions and changes to following 10 sections of the WFCR: 2.1.3.1 Cladding Degradation; 2.1.3.2 UO2 Oxidation in Fuel; 2.1.3.5 Dissolution Release from UO{sub 2}; 2.2.1.5 Fracture /Fragmentation Studies of Glass; 2.2.2.2 Dissolution Radionuclide Release from Glass; 2.2.2.3 Soluble-Precipitated/Colloidal Species from Glass; 3.2.2 Spent-Fuel Oxidation Models; 3.4.2 Spent-Fuel Dissolution Models; 3.5.1 Glass Dissolution Experimental Parameters; and 3.5.2 Glass Dissolution Models.

  17. Microscopic characterization of crystalline phases in waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Buck, E.C.; Dietz, N.L.; Wronkiewicz, D.J.; Bates, J.K. [Argonne National Lab., IL (United States); Millar, A. [Purdue Univ., West Lafayette, IN (United States)

    1995-07-01

    Transmission electron microscopy (TEM) has been used to determine the microstructure of crystalline phases present in zirconium- and titanium-bearing glass crystalline composite (GCC) waste forms. The GCC materials were found to contain spinels (maghemite), zirconolites, perovskites (CaTiO{sub 3}) and plagiociase feldspar (anorthite) mineral phases. The structure of the uranium and cerium-bearing monoclinic zirconolite was characterized by medium resolution TEM imaging and electron and X-ray diffraction (XRD). The phase was found to contain high levels of iron in comparison to Synroc-type zirconolites. Excess zirconium in zirconolite has resulted in martensitic baddeleyite (ZrO{sub 2}) formation. Anorthite (CaAl{sub 2}Si{sub 2}O{sub 8}) was present as elongated crystallites within a calcium-rich aluminosilicate glass. Lead and iron-bearing anorthite lying along distinct precipitates were occasionally observed within the an crystallographic planes.

  18. Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview

    Energy Technology Data Exchange (ETDEWEB)

    Dayal, R.; Lee, B.S.; Wilke, R.J.; Swyler, K.J.; Soo, P.; Ahn, T.M.; McIntyre, N.S.; Veakis, E.

    1982-02-01

    In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified.

  19. Material Recovery and Waste Form Development FY 2014 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Braase, Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-11-01

    Develop advanced nuclear fuel cycle separation and waste management technologies that improve current fuel cycle performance and enable a sustainable fuel cycle, with minimal processing, waste generation, and potential for material diversion.

  20. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Singh, D. [Argonne National Lab., IL (United States)

    1996-10-01

    The goal of this project is to develop and demonstrate the feasibility of a novel low-temperature solidification/stabilization (S/S) technology for immobilizing waste streams containing fission products such as cesium, strontium, and technetium in a chemically bonded phosphate ceramic. This technology can immobilize partitioned tank wastes and decontaminate waste streams containing volatile fission products.

  1. MINERALIZATION OF RADIOACTIVE WASTES BY FLUIDIZED BED STEAM REFORMING (FBSR): COMPARISONS TO VITREOUS WASTE FORMS, AND PERTINENT DURABILITY TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C

    2008-12-26

    The Savannah River National Laboratory (SRNL) was requested to generate a document for the Washington State Department of Ecology and the U.S. Environmental Protection Agency that would cover the following topics: (1) A description of the mineral structures produced by Fluidized Bed Steam Reforming (FBSR) of Hanford type Low Activity Waste (LAW including LAWR which is LAW melter recycle waste) waste, especially the cage structured minerals and how they are formed. (2) How the cage structured minerals contain some contaminants, while others become part of the mineral structure (Note that all contaminants become part of the mineral structure and this will be described in the subsequent sections of this report). (3) Possible contaminant release mechanisms from the mineral structures. (4) Appropriate analyses to evaluate these release mechanisms. (5) Why the appropriate analyses are comparable to the existing Hanford glass dataset. In order to discuss the mineral structures and how they bond contaminants a brief description of the structures of both mineral (ceramic) and vitreous waste forms will be given to show their similarities. By demonstrating the similarities of mineral and vitreous waste forms on atomic level, the contaminant release mechanisms of the crystalline (mineral) and amorphous (glass) waste forms can be compared. This will then logically lead to the discussion of why many of the analyses used to evaluate vitreous waste forms and glass-ceramics (also known as glass composite materials) are appropriate for determining the release mechanisms of LAW/LAWR mineral waste forms and how the durability data on LAW/LAWR mineral waste forms relate to the durability data for LAW/LAWR glasses. The text will discuss the LAW mineral waste form made by FBSR. The nanoscale mechanism by which the minerals form will be also be described in the text. The appropriate analyses to evaluate contaminant release mechanisms will be discussed, as will the FBSR test results to

  2. Naturally occurring glasses: analogues for radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, R.C.; Haaker, R.F.

    1979-04-01

    Volcanic glasses are very often altered by weathering and leaching and recrystallize to their fine-grained equivalents (rhyolites, felsites). The oldest volcanic glasses are dated at 40 million years before the present, but the majority are much younger. Devitrification textures was produced experimentally; and hydration rates for volcanic glasses were determined as a function of composition, temperature, and climate. Presence of water and temperature are the most important rate controlling variables. Even material that may still be described as glassy often exhibits evidence of alteration and recrystallization. Of the volcanic glasses that are preserved in the geologic record, it would be rare to describe such a glass as pristine. Despite the common alteration and recrystallization effects observed in volcanic glasses, glasses formed as a result of impact, tektites and lunar glasses, may occur in substantially unaltered form. In the case of tektites, their resistance to alteration is a result of their high SiO/sub 2/ content and low alkali content. Lunar glasses have been preserved for hundreds of millions of years because they exist in an environment with a low oxygen fugacity and an extremely low water vapor partial presssure. Thus one might expect glasses of particular compositions or in specific types of environment to be stable for long periods of time. These conclusions are applied to radioactive waste disposal over several time periods (0-30h, 30h-20y, 20-200y).

  3. Naturally occurring glasses: analogues for radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, R.C.; Haaker, R.F.

    1979-04-01

    Volcanic glasses are very often altered by weathering and leaching and recrystallize to their fine-grained equivalents (rhyolites, felsites). The oldest volcanic glasses are dated at 40 million years before the present, but the majority are much younger. Devitrification textures was produced experimentally; and hydration rates for volcanic glasses were determined as a function of composition, temperature, and climate. Presence of water and temperature are the most important rate controlling variables. Even material that may still be described as glassy often exhibits evidence of alteration and recrystallization. Of the volcanic glasses that are preserved in the geologic record, it would be rare to describe such a glass as pristine. Despite the common alteration and recrystallization effects observed in volcanic glasses, glasses formed as a result of impact, tektites and lunar glasses, may occur in substantially unaltered form. In the case of tektites, their resistance to alteration is a result of their high SiO/sub 2/ content and low alkali content. Lunar glasses have been preserved for hundreds of millions of years because they exist in an environment with a low oxygen fugacity and an extremely low water vapor partial presssure. Thus one might expect glasses of particular compositions or in specific types of environment to be stable for long periods of time. These conclusions are applied to radioactive waste disposal over several time periods (0-30h, 30h-20y, 20-200y).

  4. Determination of the Rate of Formation of Hydroceramic Waste Forms made with INEEL Calcined Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Barry Scheetz; Johnson Olanrewaju

    2001-10-15

    The formulation, synthesis, characterization and hydration kinetics of hydroceramic waste forms designed as potential hosts for existing INEEL calcine high-level wastes have been established as functions of temperature and processing time. Initial experimentations were conducted with several aluminosilicate pozzolanic materials, ranging from fly ash obtained from various power generating coal and other combustion industries to reactive alumina, natural clays and ground bottled glass powders. The final selection criteria were based on the ease of processing, excellent physical properties and chemical durability (low-leaching) determined from the PCT test produced in hydroceramic. The formulation contains vermiculite, Sr(NO32), CsC1, NaOH, thermally altered (calcined natural clay) and INEEL simulated calcine high-level nuclear wastes and 30 weight percent of fluorinel blend calcine and zirconia calcine. Syntheses were carried out at 75-200 degree C at autogeneous water pressure (100% relative humidity) at various time intervals. The resulting monolithic compact products were hard and resisted breaking when dropped from a 5 ft height. Hydroceramic host mixed with fluorinel blend calcine and processed at 75 degree C crumbled into rice hull-side grains or developed scaly flakes. However, the samples equally possessed the same chemical durability as their unbroken counterparts. Phase identification by XRD revealed that hydroceramic host crystallized type zeolite at 75-150 degree C and NaP1 at 175-200 degree C in addition to the presence of quartz phase originating from the clay reactant. Hydroceramic host mixed with either fluorinel blend calcine or zirconia calcine crystallized type A zeolite at 75-95 degree C, formed a mixture of type A zeolite and hydroxysodalite at 125-150 degree C and hydroxysodalite at 175-200 degree C. Quartz, calcium fluoride and zirconia phases from the clay reactant and the two calcine wastes were also detected. The PCT test solution

  5. Annual report on the development and characterization of solidified forms for nuclear wastes, 1979

    Energy Technology Data Exchange (ETDEWEB)

    Chick, L.A.; McVay, G.L.; Mellinger, G.B.; Roberts, F.P.

    1980-12-01

    Development and characterization of solidified nuclear waste forms is a major continuing effort at Pacific Northwest Laboratory. Contributions from seven programs directed at understanding chemical composition, process conditions, and long-term behaviors of various nuclear waste forms are included in this report. The major findings of the report are included in extended figure captions that can be read as brief technical summaries of the research, with additional information included in a traditional narrative format. Waste form development proceeded on crystalline and glass materials for high-level and transuranic (TRU) wastes. Leaching studies emphasized new areas of research aimed at more basic understanding of waste form/aqueous solution interactions. Phase behavior and thermal effects research included studies on crystal phases in defense and TRU waste glasses and on liquid-liquid phase separation in borosilicate waste glasses. Radiation damage effects in crystals and glasses from alpha decay and from transmutation are reported.

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

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

  7. Nuclear waste form risk assessment for US defense waste at Savannah River Plant. Annual report fiscal year 1980

    Energy Technology Data Exchange (ETDEWEB)

    Cheung, H.; Jackson, D.D.; Revelli, M.A.

    1981-07-01

    Waste form dissolution studies and preliminary performance analyses were carried out to contribute a part of the data needed for the selection of a waste form for the disposal of Savannah River Plant defense waste in a deep geologic repository. The first portion of this work provides descriptions of the chemical interactions between the waste form and the geologic environment. We reviewed critically the dissolution/leaching data for borosilicate glass and SYNROC. Both chemical kinetic and thermodynamic models were developed to describe the dissolution process of these candidate waste forms so as to establish a fundamental basis for interpretation of experimental data and to provide directions for future experiments. The complementary second portion of this work is an assessment of the impacts of alternate waste forms upon the consequences of disposal in various proposed geological media. Employing systems analysis methodology, we began to evaluate the performance of a generic waste form for the case of a high risk scenario for a bedded salt repository. Results of sensitivity analysis, uncertainty analyses, and sensitivity to uncertainty analysis are presented.

  8. 40 CFR 761.205 - Notification of PCB waste activity (EPA Form 7710-53).

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Notification of PCB waste activity..., DISTRIBUTION IN COMMERCE, AND USE PROHIBITIONS PCB Waste Disposal Records and Reports § 761.205 Notification of PCB waste activity (EPA Form 7710-53). (a)(1) All commercial storers, transporters, and disposers...

  9. Secondary Waste Form Development and Optimization—Cast Stone

    Energy Technology Data Exchange (ETDEWEB)

    Sundaram, S. K.; Parker, Kent E.; Valenta, Michelle M.; Pitman, Stan G.; Chun, Jaehun; Chung, Chul-Woo; Kimura, Marcia L.; Burns, Carolyn A.; Um, Wooyong; Westsik, Joseph H.

    2011-07-14

    Washington River Protection Services is considering the design and construction of a Solidification Treatment Unit (STU) for the Effluent Treatment Facility (ETF) at Hanford. The ETF is a Resource Conservation and Recovery Act-permitted, multi-waste, treatment and storage unit and can accept dangerous, low-level, and mixed wastewaters for treatment. The STU needs to be operational by 2018 to receive secondary liquid wastes generated during operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The STU to ETF will provide the additional capacity needed for ETF to process the increased volume of secondary wastes expected to be produced by WTP.

  10. NNWSI [Nevada Nuclear Waste Storage Investigations] waste form testing at Argonne National Laboratory; Semiannual report, January--June 1988

    Energy Technology Data Exchange (ETDEWEB)

    Bates, J.K.; Gerding, T.J.; Ebert, W.L.; Mazer, J.J.; Biwer, B.M. [Argonne National Lab., IL (USA)

    1990-04-01

    The Chemical Technology Division of Argonne National Laboratory is performing experiments in support of the waste package development of the Yucca Mountain Project (formerly the Nevada Nuclear Waste Storage Investigations Project). Experiments in progress include (1) the development and performance of a durability test in unsaturated conditions, (2) studies of waste form behavior in an irradiated atmosphere, (3) studies of behavior in water vapor, and (4) studies of naturally occurring glasses to be used as analogues for waste glass behavior. This report documents progress made during the period of January--June 1988. 21 refs., 37 figs., 12 tabs.

  11. Durability and degradation of HT9 based alloy waste forms with variable Ni and Cr content

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-31

    Short-term electrochemical and long-term hybrid electrochemical corrosion tests were performed on alloy waste forms in reference aqueous solutions that bound postulated repository conditions. The alloy waste forms investigated represent candidate formulations that can be produced with advanced electrochemical treatment of used nuclear fuel. The studies helped to better understand the alloy waste form durability with differing concentrations of nickel and chromium, species that can be added to alloy waste forms to potentially increase their durability and decrease radionuclide release into the environment.

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

    OpenAIRE

    Carmalin Sophia Ayyappan

    2015-01-01

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

  13. Spent fuel treatment and mineral waste form development at Argonne National Laboratory-West

    Energy Technology Data Exchange (ETDEWEB)

    Goff, K.M.; Benedict, R.W.; Bateman, K. [Argonne National Lab., Idaho Falls, ID (United States); Lewis, M.A.; Pereira, C. [Argonne National Lab., IL (United States); Musick, C.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1996-07-01

    At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. Both mineral and metal high-level waste forms will be produced. The mineral waste form will contain the active metal fission products and the transuranics. Cold small-scale waste form testing has been on-going at Argonne in Illinois. Large-scale testing is commencing at ANL-West.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-03-01

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

  15. Transuranic contaminated waste form characterization and data base

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-07-01

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

  16. Description of Defense Waste Processing Facility reference waste form and canister. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, R.G.

    1983-08-01

    The Defense Waste Processing Facility (DWPF) will be located at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1984. The reference waste form is borosilicate glass containing approx. 28 wt % sludge oxides, with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains about 58% SiO/sub 2/ and 15% B/sub 2/O/sub 3/. Leachabilities of SRP waste glasses are expected to approach 10/sup -8/ g/m/sup 2/-day based upon 1000-day tests using glasses containing SRP radioactive waste. Tests were performed under a wide variety of conditions simulating repository environments. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approx. 470 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the sludge and supernate processes. The radionuclide content of the canister is about 177,000 ci, with a radiation level of 5500 rem/h at canister surface contact. The reference canister is fabricated of standard 24-in.-OD, Schedule 20, 304L stainless steel pipe with a dished bottom, domed head, and a combined lifting and welding flange on the head neck. The overall canister length is 9 ft 10 in. with a 3/8-in. wall thickness. The 3-m canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected as an optimum size from glass quality considerations, a logical size for repository handling and to ensure that a filled canister with its double containment shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be compatible with preliminary assessments of repository requirements. 10 references.

  17. Molecular Environmental Science Using Synchrotron Radiation: Chemistry and Physics of Waste Form Materials

    Energy Technology Data Exchange (ETDEWEB)

    Lindle, Dennis W.

    2011-04-21

    Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization. Specially formulated glass compositions and ceramics such as pyrochlores and apatites are the main candidates for these wastes. An important consideration linked to the durability of waste-form materials is the local structure around the waste components. Equally important is the local structure of constituents of the glass and ceramic host matrix. Knowledge of the structure in the waste-form host matrices is essential, prior to and subsequent to waste incorporation, to evaluate and develop improved waste-form compositions based on scientific considerations. This project used the soft-x-ray synchrotron-radiation-based technique of near-edge x-ray-absorption fine structure (NEXAFS) as a unique method for investigating oxidation states and structures of low-Z elemental constituents forming the backbones of glass and ceramic host matrices for waste-form materials. In addition, light metal ions in ceramic hosts, such as titanium, are also ideal for investigation by NEXAFS in the soft-x-ray region. Thus, one of the main objectives was to understand outstanding issues in waste-form science via NEXAFS investigations and to translate this understanding into better waste-form materials, followed by eventual capability to investigate “real” waste-form materials by the same methodology. We conducted several detailed structural investigations of both pyrochlore ceramic and borosilicate-glass materials during the project and developed improved capabilities at Beamline 6.3.1 of the Advanced Light Source (ALS) to perform the studies.

  18. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    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 multi-phase 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. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  19. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    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 multi-phase 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. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

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

  1. Heat of Hydration of Low Activity Cementitious Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-23

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

  2. Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

    2011-12-01

    This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

  3. Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

    Energy Technology Data Exchange (ETDEWEB)

    1983-06-01

    The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available.

  4. Mineral assemblage transformation of a metakaolin-based waste form after geopolymer encapsulation

    Science.gov (United States)

    Williams, Benjamin D.; Neeway, James J.; Snyder, Michelle M. V.; Bowden, Mark E.; Amonette, James E.; Arey, Bruce W.; Pierce, Eric M.; Brown, Christopher F.; Qafoku, Nikolla P.

    2016-05-01

    Mitigation of hazardous and radioactive waste can be improved through conversion of existing waste to a more chemically stable and physically robust waste form. One option for waste conversion is the fluidized bed steam reforming (FBSR) process. The resulting FBSR granular material was encapsulated in a geopolymer matrix referred to here as Geo-7. This provides mechanical strength for ease in transport and disposal. However, it is necessary to understand the phase assemblage evolution as a result of geopolymer encapsulation. In this study, we examine the mineral assemblages formed during the synthesis of the multiphase ceramic waste form. The FBSR granular samples were created from waste simulant that was chemically adjusted to resemble Hanford tank waste. Another set of samples was created using Savannah River Site Tank 50 waste simulant in order to mimic a blend of waste collected from 68 Hanford tank. Waste form performance tests were conducted using the product consistency test (PCT), the Toxicity Characteristic Leaching Procedure (TCLP), and the single-pass flow-through (SPFT) test. X-ray diffraction analyses revealed the structure of a previously unreported NAS phase and indicate that monolith creation may lead to a reduction in crystallinity as compared to the primary FBSR granular product.

  5. Advanced waste form and Melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both

  6. In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    H.W> Stockman; S. LeStrange

    2000-09-28

    The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

  7. A study on characterization and evaluation methodologies of radioactive waste forms for safe disposal

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Y. C.; Lee, G. S.; Kim, G. J.; Nam, H.; Seok, J. H. [Yonsei Univ., Seoul (Korea, Republic of)

    2004-02-15

    The contents and scope of the study are summarized as follows : elicitation of significant items for characteristic assessment about stability analysis of radioactive waste forms for safe disposal, compressive strength, free water, leaching rate, and weatherability. Suggestion of assessment methods through the characteristic test of waste forms, comparison of assessment methods and suggestion of suitable testing methods about the above stated 4 items. Assessment modeling development for long-term stability of radioactive waste forms, weatherometric test of waste forms, expectation modeling development through VOM(Valance-Oxygen Model). Suggestion of determination standard together assessment testing methods and description about the standard. Explanation to be suitable guideline and regulation of waste handling and acceptance.

  8. Data Package for Secondary Waste Form Down-Selection—Cast Stone

    Energy Technology Data Exchange (ETDEWEB)

    Serne, R. Jeffrey; Westsik, Joseph H.

    2011-09-05

    Available literature on Cast Stone and Saltstone was reviewed with an emphasis on determining how Cast Stone and related grout waste forms performed in relationship to various criteria that will be used to decide whether a specific type of waste form meets acceptance criteria for disposal in the Integrated Disposal Facility (IDF) at Hanford. After the critical review of the Cast Stone/Saltstone literature, we conclude that Cast Stone is a good candidate waste form for further consideration. Cast stone meets the target IDF acceptance criteria for compressive strength, no free liquids, TCLP leachate are below the UTS permissible concentrations and leach rates for Na and Tc-99 are suiteably low. The cost of starting ingredients and equipment necessary to generate Cast Stone waste forms with secondary waste streams are low and the Cast Stone dry blend formulation can be tailored to accommodate variations in liquid waste stream compositions. The database for Cast Stone short-term performance is quite extensive compared to the other three candidate waste solidification processes. The solidification of liquid wastes in Cast Stone is a mature process in comparison to the other three candidates. Successful production of Cast Stone or Saltstone has been demonstrated from lab-scale monoliths with volumes of cm3 through m3 sized blocks to 210-liter sized drums all the way to the large pours into vaults at Savannah River. To date over 9 million gallons of low activity liquid waste has been solidified and disposed in concrete vaults at Savannah River.

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

    Directory of Open Access Journals (Sweden)

    Carmalin Sophia Ayyappan

    2015-06-01

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

  10. Advanced waste form and melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-02

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these "troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approached to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.

  11. FY16 Annual Accomplishments - Waste Form Development and Performance: Evaluation Of Ceramic Waste Forms - Comparison Of Hot Isostatic Pressed And Melt Processed Fabrication Methods

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dandeneau, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-10-13

    FY16 efforts were focused on direct comparison of multi-phase ceramic waste forms produced via melt processing and HIP methods. Based on promising waste form compositions previously devised at SRNL, simulant material was prepared at SRNL and a portion was sent to the Australian Nuclear Science and Technology Organization (ANSTO) for HIP treatments, while the remainder of the material was melt processed at SRNL. The microstructure, phase formation, elemental speciation, and leach behavior, and radiation stability of the fabricated ceramics was performed. In addition, melt-processed ceramics designed with different fractions of hollandite, zirconolite, perovskite, and pyrochlore phases were investigated. for performance and properties.

  12. Molecular environmental science using synchrotron radiation:Chemistry and physics of waste form materials

    Energy Technology Data Exchange (ETDEWEB)

    Lindle, Dennis W.; Shuh, David K.

    2005-02-28

    Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization [1]. Specially formulated glass compositions, many of which have been derived from glass developed for commercial purposes, and ceramics such as pyrochlores and apatites, will be the main recipients for these wastes. The performance characteristics of waste-form glasses and ceramics are largely determined by the loading capacity for the waste constituents (radioactive and non-radioactive) and the resultant chemical and radiation resistance of the waste-form package to leaching (durability). There are unique opportunities for the use of near-edge soft-x-ray absorption fine structure (NEXAFS) spectroscopy to investigate speciation of low-Z elements forming the backbone of waste-form glasses and ceramics. Although nuclear magnetic resonance (NMR) is the primary technique employed to obtain speciation information from low-Z elements in waste forms, NMR is incompatible with the metallic impurities contained in real waste and is thus limited to studies of idealized model systems. In contrast, NEXAFS can yield element-specific speciation information from glass constituents without sensitivity to paramagnetic species. Development and use of NEXAFS for eventual studies of real waste glasses has significant implications, especially for the low-Z elements comprising glass matrices [5-7]. The NEXAFS measurements were performed at Beamline 6.3.1, an entrance-slitless bend-magnet beamline operating from 200 eV to 2000 eV with a Hettrick-Underwood varied-line-space (VLS) grating monochromator, of the Advanced Light Source (ALS) at LBNL. Complete characterization and optimization of this beamline was conducted to enable high-performance measurements.

  13. U.S. Food Loss and Waste 2030 Champions Activity Form

    Science.gov (United States)

    To join the U.S. Food Loss and Waste 2030 Champions, organizations complete and submit the 2030 Champions form, in which they commit to reduce food loss and waste in their own operations and periodically report their progress on their website.

  14. A Science-based Approach to Development of Durable Waste Forms

    Science.gov (United States)

    Peters, M. T.; Ewing, R. C.

    2006-05-01

    There are two compelling reasons for the importance of understanding the source term and near-field processes in a geologic repository. First, almost all of the radioactivity is initially in the waste form, mainly in the spent nuclear fuel (SNF) or nuclear waste glass. Second, over long periods, after the engineered barriers are no longer important, it is the waste form that controls the release of radioactivity. Thus, it is essential to know the physical and chemical state of the waste form after hundreds of thousands of years. The United States Department of Energy's Yucca Mountain Repository Program has initiated a long-term program to develop a basic understanding of the fundamental mechanisms of radionuclide release and a quantification of the release as repository conditions evolve over time. Specifically, the research program addresses four critical areas: a) SNF dissolution mechanisms and rates; b) formation and properties of U6+- secondary phases; c) waste form-waste package interactions in the near-field; and d) integration of in-package chemical and physical processes. The ultimate goal is to integrate the scientific results into a larger scale model of the source term and near-field processes. This integrated model will be used to provide a basis for understanding the behavior of the source term over long time periods (greater than 100,000 years). Such a fundamental and integrated experimental and modeling approach to source term processes can also be readily applied to development of advanced waste forms as part of a closed nuclear fuel cycle. Specifically, a fundamental understanding of candidate waste form materials stability in high temperature/high radiation environments and near-field geochemical/hydrologic processes could enable development of advanced waste forms "tailored" to specific geologic settings.

  15. Glass binder development for a glass-bonded sodalite ceramic waste form

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Vienna, John D.; Frank, Steven M.; Kroll, Jared O.; Peterson, Jacob A.; Canfield, Nathan L.; Zhu, Zihua; Zhang, Jiandong; Kruska, Karen; Schreiber, Daniel K.; Crum, Jarrod V.

    2017-06-01

    This paper discusses work to develop Na2O-B2O3-SiO2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. Here, five new glasses with high Na2O contents were designed to generate waste forms having higher sodalite contents and fewer stress fractures. The structural, mechanical, and thermal properties of the new glasses were measured using variety of analytical techniques. The glasses were then used to produce ceramic waste forms with surrogate salt waste. The materials made using the glasses developed during this study were formulated to generate more sodalite than materials made with previous baseline glasses used. The coefficients of thermal expansion for the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature. These improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability. Additionally, a model generated during this study for predicting softening temperature of silicate binder glasses is presented.

  16. Summary of Uranium Solubility Studies in Concrete Waste Forms and Vadose Zone Environments

    Energy Technology Data Exchange (ETDEWEB)

    Golovich, Elizabeth C.; Wellman, Dawn M.; Serne, R. Jeffrey; Bovaird, Chase C.

    2011-09-30

    One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Concrete encasement would contain and isolate the waste packages from the hydrologic environment and act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. This report presents the results of investigations elucidating the uranium mineral phases controlling the long-term fate of uranium within concrete waste forms and the solubility of these phases in concrete pore waters and alkaline, circum-neutral vadose zone environments.

  17. Chemical and Charge Imbalance Induced by Radionuclide Decay: Effects on Waste Form Structure

    Energy Technology Data Exchange (ETDEWEB)

    Van Ginhoven, Renee M.; Jaffe, John E.; Jiang, Weilin; Strachan, Denis M.

    2011-04-01

    This is a milestone document covering the activities to validate theoretical calculations with experimental data for the effect of the decay of 90Sr to 90Zr on materials properties. This was done for a surragate waste form strontium titanate.

  18. Coupling of Nuclear Waste Form Corrosion and Radionuclide Transports in Presence of Relevant Repository Sediments

    Energy Technology Data Exchange (ETDEWEB)

    Wall, Nathalie A. [Washington State Univ., Pullman, WA (United States); Neeway, James J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qafoku, Nikolla P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ryan, Joseph V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-09-30

    Assessments of waste form and disposal options start with the degradation of the waste forms and consequent mobilization of radionuclides. Long-term static tests, single-pass flow-through tests, and the pressurized unsaturated flow test are often employed to study the durability of potential waste forms and to help create models that predict their durability throughout the lifespan of the disposal site. These tests involve the corrosion of the material in the presence of various leachants, with different experimental designs yielding desired information about the behavior of the material. Though these tests have proved instrumental in elucidating various mechanisms responsible for material corrosion, the chemical environment to which the material is subject is often not representative of a potential radioactive waste repository where factors such as pH and leachant composition will be controlled by the near-field environment. Near-field materials include, but are not limited to, the original engineered barriers, their resulting corrosion products, backfill materials, and the natural host rock. For an accurate performance assessment of a nuclear waste repository, realistic waste corrosion experimental data ought to be modeled to allow for a better understanding of waste form corrosion mechanisms and the effect of immediate geochemical environment on these mechanisms. Additionally, the migration of radionuclides in the resulting chemical environment during and after waste form corrosion must be quantified and mechanisms responsible for migrations understood. The goal of this research was to understand the mechanisms responsible for waste form corrosion in the presence of relevant repository sediments to allow for accurate radionuclide migration quantifications. The rationale for this work is that a better understanding of waste form corrosion in relevant systems will enable increased reliance on waste form performance in repository environments and potentially

  19. Advanced waste forms research and development. First quarterly report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, G.J.

    1975-08-05

    Activities during the last two months are described. A significant portion of time was spent reviewing the literature on the Cs/sub 2/O-Al/sub 2/O/sub 3/-SiO/sub 2/ system, on the use of clays and zeolites for Cs-fixation of aqueous wastes, and on silicate-phosphate apatite structure crystal chemistry. The results from the latest group of compatibility studies (CS-runs) were used to modify the first demonstration supercalcine formulation so that it is more in line with the actual crystalline phase formation. Supercalcine formuation 75-2 is described.

  20. Preliminary evaluation of alternative forms for immobilization of Savannah River Plant high-level waste. [Eleven alternative solid forms

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Comparison of mechanical properties of glass-bonded sodalite and borosilicate glass high-level waste forms

    Energy Technology Data Exchange (ETDEWEB)

    O' Holleran, T. P.; DiSanto, T.; Johnson, S. G.; Goff, K. M.

    2000-05-09

    Argonne National Laboratory has developed a glass-bonded sodalite waste form to immobilize the salt waste stream from electrometallurgical treatment of spent nuclear fuel. The waste form consists of 75 vol.% crystalline sodalite and 25 vol.% glass. Microindentation fracture toughness measurements were performed on this material and borosilicate glass from the Defense Waste Processing Facility using a Vickers indenter. Palmqvist cracking was confined for the glass-bonded sodalite waste form, while median-radial cracking occurred in the borosilicate glass. The elastic modulus was measured by an acoustic technique. Fracture toughness, microhardness, and elastic modulus values are reported for both waste forms.

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

    Energy Technology Data Exchange (ETDEWEB)

    SAMS TL; EDGE JA; SWANBERG DJ; ROBBINS RA

    2011-01-13

    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.

  3. Far-Field Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    J.P. Nicot

    2000-09-29

    The objective of this calculation is to estimate the quantity of fissile material that could accumulate in fractures in the rock beneath plutonium-ceramic (Pu-ceramic) and Mixed-Oxide (MOX) waste packages (WPs) as they degrade in the potential monitored geologic repository at Yucca Mountain. This calculation is to feed another calculation (Ref. 31) computing the probability of criticality in the systems described in Section 6 and then ultimately to a more general report on the impact of plutonium on the performance of the proposed repository (Ref. 32), both developed concurrently to this work. This calculation is done in accordance with the development plan TDP-DDC-MD-000001 (Ref. 9), item 5. The original document described in item 5 has been split into two documents: this calculation and Ref. 4. The scope of the calculation is limited to only very low flow rates because they lead to the most conservative cases for Pu accumulation and more generally are consistent with the way the effluent from the WP (called source term in this calculation) was calculated (Ref. 4). Ref. 4 (''In-Drift Accumulation of Fissile Material from WPs Containing Plutonium Disposition Waste Forms'') details the evolution through time (breach time is initial time) of the chemical composition of the solution inside the WP as degradation of the fuel and other materials proceed. It is the chemical solution used as a source term in this calculation. Ref. 4 takes that same source term and reacts it with the invert; this calculation reacts it with the rock. In addition to reactions with the rock minerals (that release Si and Ca), the basic mechanisms for actinide precipitation are dilution and mixing with resident water as explained in Section 2.1.4. No other potential mechanism such as flow through a reducing zone is investigated in this calculation. No attempt was made to use the effluent water from the bottom of the invert instead of using directly the effluent water from the

  4. RADIOACTIVE DEMONSTRATION OF FINAL MINERALIZED WASTE FORMS FOR HANFORD WASTE TREATMENT PLANT SECONDARY WASTE BY FLUIDIZED BED STEAM REFORMING USING THE BENCH SCALE REFORMER PLATFORM

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C.; Burket, P.; Cozzi, A.; Daniel, W.; Jantzen, C.; Missimer, D.

    2012-02-02

    ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage, but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the SRNL to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. BSR testing with WTP SW waste surrogates and associated analytical analyses and tests of granular products (GP) and monoliths began in the Fall of 2009, and then was continued from the Fall of 2010 through the Spring of 2011. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of {sup 125/129}I and {sup 99}Tc to chemically resemble WTP-SW. Prior to these radioactive feed tests, non-radioactive simulants were also processed. Ninety six grams of radioactive granular product were made for testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.

  5. Performance of a Steel/Oxide Composite Waste Form for Combined Waste Steams from Advanced Electrochemical Processes

    Energy Technology Data Exchange (ETDEWEB)

    Indacochea, J. E. [Univ. of Illinois, Chicago, IL (United States); Gattu, V. K. [Univ. of Illinois, Chicago, IL (United States); Chen, X. [Univ. of Illinois, Chicago, IL (United States); Rahman, T. [Univ. of Illinois, Chicago, IL (United States)

    2017-06-15

    The results of electrochemical corrosion tests and modeling activities performed collaboratively by researchers at the University of Illinois at Chicago and Argonne National Laboratory as part of workpackage NU-13-IL-UIC-0203-02 are summarized herein. The overall objective of the project was to develop and demonstrate testing and modeling approaches that could be used to evaluate the use of composite alloy/ceramic materials as high-level durable waste forms. Several prototypical composite waste form materials were made from stainless steels representing fuel cladding, reagent metals representing metallic fuel waste streams, and reagent oxides representing oxide fuel waste streams to study the microstructures and corrosion behaviors of the oxide and alloy phases. Microelectrodes fabricated from small specimens of the composite materials were used in a series of electrochemical tests to assess the corrosion behaviors of the constituent phases and phase boundaries in an aggressive acid brine solution at various imposed surface potentials. The microstructures were characterized in detail before and after the electrochemical tests to relate the electrochemical responses to changes in both the electrode surface and the solution composition. The results of microscopic, electrochemical, and solution analyses were used to develop equivalent circuit and physical models representing the measured corrosion behaviors of the different materials pertinent to long-term corrosion behavior. This report provides details regarding (1) the production of the composite materials, (2) the protocol for the electrochemical measurements and interpretations of the responses of multi-phase alloy and oxide composites, (3) relating corrosion behaviors to microstructures of multi-phase alloys based on 316L stainless steel and HT9 (410 stainless steel was used as a substitute) with added Mo, Ni, and/or Mn, and (4) modeling the corrosion behaviors and rates of several alloy/oxide composite

  6. Leaching tests of simulated Cogema bituminized waste form

    Energy Technology Data Exchange (ETDEWEB)

    Nakayama, S.; Akimoto, T.; Iida, Y.; Nagano, T. [Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-07-01

    The leaching behavior of COGEMA-type bituminized radioactive waste was studied for the atmospheric and anaerobic conditions. Active and inactive laboratory-scale bitumen samples, including two major salts of NaNO{sub 3} and BaSO{sub 4}, were contacted with deionized water, an alkaline solution (0.01 mol/L Ca(OH){sub 2} or 0.03 mol/L KOH), or a saline solution (0.5 mol/L KCl). It was found that the release of salt was reduced in the Ca(OH){sub 2} solution compared with deionized water under the atmospheric conditions. No significant difference in the concentrations of {sup 237}Np in leachants contacted with the samples for 7 days was observed between the atmospheric and the anaerobic conditions. (authors)

  7. Transuranic and Low-Level Boxed Waste Form Nondestructive Assay Technology Overview and Assessment

    Energy Technology Data Exchange (ETDEWEB)

    G. Becker; M. Connolly; M. McIlwain

    1999-02-01

    The Mixed Waste Focus Area (MWFA) identified the need to perform an assessment of the functionality and performance of existing nondestructive assay (NDA) techniques relative to the low-level and transuranic waste inventory packaged in large-volume box-type containers. The primary objectives of this assessment were to: (1) determine the capability of existing boxed waste form NDA technology to comply with applicable waste radiological characterization requirements, (2) determine deficiencies associated with existing boxed waste assay technology implementation strategies, and (3) recommend a path forward for future technology development activities, if required. Based on this assessment, it is recommended that a boxed waste NDA development and demonstration project that expands the existing boxed waste NDA capability to accommodate the indicated deficiency set be implemented. To ensure that technology will be commercially available in a timely fashion, it is recommended this development and demonstration project be directed to the private sector. It is further recommended that the box NDA technology be of an innovative design incorporating sufficient NDA modalities, e.g., passive neutron, gamma, etc., to address the majority of the boxed waste inventory. The overall design should be modular such that subsets of the overall NDA system can be combined in optimal configurations tailored to differing waste types.

  8. Preliminary Waste Form Compliance Plan for the Idaho National Engineering and Environmental Laboratory High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    B. A. Staples; T. P. O' Holleran

    1999-05-01

    The Department of Energy (DOE) has specific technical and documentation requirements for high-level waste (HLW) that is to be placed in a federal repository. This document describes in general terms the strategy to be used at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that vitrified HLW, if produced at the INEEL, meets these requirements. Waste form, canister, quality assurance, and documentation specifications are discussed. Compliance strategy is given, followed by an overview of how this strategy would be implemented for each specification.

  9. Computational Efficient Upscaling Methodology for Predicting Thermal Conductivity of Nuclear Waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Li, Dongsheng; Sun, Xin; Khaleel, Mohammad A.

    2011-09-28

    This study evaluated different upscaling methods to predict thermal conductivity in loaded nuclear waste form, a heterogeneous material system. The efficiency and accuracy of these methods were compared. Thermal conductivity in loaded nuclear waste form is an important property specific to scientific researchers, in waste form Integrated performance and safety code (IPSC). The effective thermal conductivity obtained from microstructure information and local thermal conductivity of different components is critical in predicting the life and performance of waste form during storage. How the heat generated during storage is directly related to thermal conductivity, which in turn determining the mechanical deformation behavior, corrosion resistance and aging performance. Several methods, including the Taylor model, Sachs model, self-consistent model, and statistical upscaling models were developed and implemented. Due to the absence of experimental data, prediction results from finite element method (FEM) were used as reference to determine the accuracy of different upscaling models. Micrographs from different loading of nuclear waste were used in the prediction of thermal conductivity. Prediction results demonstrated that in term of efficiency, boundary models (Taylor and Sachs model) are better than self consistent model, statistical upscaling method and FEM. Balancing the computation resource and accuracy, statistical upscaling is a computational efficient method in predicting effective thermal conductivity for nuclear waste form.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bleier, A.

    1997-09-01

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

  11. I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING

    Energy Technology Data Exchange (ETDEWEB)

    S.M. Frank

    2011-09-01

    Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomic Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this

  12. Determination of the Structure of Vitrified Hydroceramic/CBC Waste Form Glasses Manufactured from DOE Reprocessing Waste

    Energy Technology Data Exchange (ETDEWEB)

    Scheetz, B.E.; White, W. B.; Chesleigh, M.; Portanova, A.; Olanrewaju, J.

    2005-05-31

    The selection of a glass-making option for the solidification of nuclear waste has dominated DOE waste form programs since the early 1980's. Both West Valley and Savannah River are routinely manufacturing glass logs from the high level waste inventory in tank sludges. However, for some wastes, direct conversion to glass is clearly not the optimum strategy for immobilization. INEEL, for example, has approximately 4400 m{sup 3} of calcined high level waste with an activity that produces approximately 45 watts/m{sup 3}, a rather low concentration of radioactive constituents. For these wastes, there is value in seeking alternatives to glass. An alternative approach has been developed and the efficacy of the process demonstrated that offers a significant savings in both human health and safety exposures and also a lower cost relative to the vitrification option. The alternative approach utilizes the intrinsic chemical reactivity of the highly alkaline waste with the addition of aluminosilicate admixtures in the appropriate proportions to form zeolites. The process is one in which a chemically bonded ceramic is produced. The driving force for reaction is derived from the chemical system itself at very modest temperatures and yet forms predominantly crystalline phases. Because the chemically bonded ceramic requires an aqueous medium to serve as a vehicle for the chemical reaction, the proposed zeolite-containing waste form can more adequately be described as a hydroceramic. The hydrated crystalline materials are then subject to hot isostatic pressing (HIP) which partially melts the material to form a glass ceramic. The scientific advantages of the hydroceramic/CBC approach are: (1) Low temperature processing; (2) High waste loading and thus only modest volumetric bulking from the addition of admixtures; (3) Ability to immobilize sodium; (4) Ability to handle low levels of nitrate (2-3% NO{sub 3}{sup -}); (5) The flexibility of a vitrifiable waste; and (6) A process

  13. Annual report Development and characterization of solidified forms for high-level wastes: 1978.

    Energy Technology Data Exchange (ETDEWEB)

    Ross, W.A.; Mendel, J.E.

    1979-12-01

    Development and characterization of solidified high-level waste forms are directed at determining both process properties and long-term behaviors of various solidified high-level waste forms in aqueous, thermal, and radiation environments. Waste glass properties measured as a function of composition were melt viscosity, melt electrical conductivity, devitrification, and chemical durability. The alkali metals were found to have the greatest effect upon glass properties. Titanium caused a slight decrease in viscosity and a significant increase in chemical durability in acidic solutions (pH-4). Aluminum, nickel and iron were all found to increase the formation of nickel-ferrite spinel crystals in the glass. Four multibarrier advanced waste forms were produced on a one-liter scale with simulated waste and characterized. Glass marbles encapsulated in a vacuum-cast lead alloy provided improved inertness with a minimal increase in technological complexity. Supercalcine spheres exhibited excellent inertness when coated with pyrolytic carbon and alumina and put in a metal matrix, but the processing requirements are quite complex. Tests on simulated and actual high-level waste glasses continue to suggest that thermal devitrification has a relatively small effect upon mechanical and chemical durabilities. Tests on the effects radiation has upon waste forms also continue to show changes to be relatively insignificant. Effects caused by decay of actinides can be estimated to saturate at near 10/sup 19/ alpha-events/cm/sup 3/ in homogeneous solids. Actually, in solidified waste forms the effects are usually observed around certain crystals as radiation causes amorphization and swelling of th crystals.

  14. Characteristics of high-level radioactive waste forms for their disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung

    2000-12-01

    In order to develop a deep geological repository for a high-level radioactive waste coming from reprocessing of spent nuclear fuels discharged from our domestic nuclear power plants, the the required characteristics of waste form are dependent upon a solidifying medium and the amount of waste loading in the medium. And so, by the comparative analysis of the characteristics of various waste forms developed up to the present, a suitable medium is recommended.The overall characteristics of the latter is much better than those of the former, but the change of the properties due to an amorphysation by radiation exposure and its thermal expansion has not been clearly identified yet. And its process has not been commercialized. However, the overall properties of the borosilicate glass waste forms are acceptable for their disposal, their production cost is reasonable and their processes have already been commercialized. And plenty informations of their characteristics and operational experiences have been accumulated. Consequently, it is recommended that a suitable medium solidifying the HLW is a borosilicate glass and its composition for the identification of a reference waste form would be based on the glass frit of R7T7.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-09-01

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

  17. Fracture toughness measurements on a glass bonded sodalite high-level waste form.

    Energy Technology Data Exchange (ETDEWEB)

    DiSanto, T.; Goff, K. M.; Johnson, S. G.; O' Holleran, T. P.

    1999-05-19

    The electrometallurgical treatment of metallic spent nuclear fuel produces two high-level waste streams; cladding hulls and chloride salt. Argonne National Laboratory is developing a glass bonded sodalite waste form to immobilize the salt waste stream. The waste form consists of 75 Vol.% crystalline sodalite (containing the salt) with 25 Vol.% of an ''intergranular'' glassy phase. Microindentation fracture toughness measurements were performed on representative samples of this material using a Vickers indenter. Palmqvist cracking was confirmed by post-indentation polishing of a test sample. Young's modulus was measured by an acoustic technique. Fracture toughness, microhardness, and Young's modulus values are reported, along with results from scanning electron microscopy studies.

  18. Radionuclide Retention Mechanisms in Secondary Waste-Form Testing: Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Um, Wooyong; Valenta, Michelle M.; Chung, Chul-Woo; Yang, Jungseok; Engelhard, Mark H.; Serne, R. Jeffrey; Parker, Kent E.; Wang, Guohui; Cantrell, Kirk J.; Westsik, Joseph H.

    2011-09-26

    This report describes the results from laboratory tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate candidate stabilization technologies that have the potential to successfully treat liquid secondary waste stream effluents produced by the Hanford Tank Waste Treatment and Immobilization Plant (WTP). WRPS is considering the design and construction of a Solidification Treatment Unit (STU) for the Effluent Treatment Facility (ETF) at Hanford. The ETF, a multi-waste, treatment-and-storage unit that has been permitted under the Resource Conservation and Recovery Act (RCRA), can accept dangerous, low-level, and mixed wastewaters for treatment. The STU needs to be operational by 2018 to receive secondary liquid waste generated during operation of the WTP. The STU will provide the additional capacity needed for ETF to process the increased volume of secondary waste expected to be produced by WTP. This report on radionuclide retention mechanisms describes the testing and characterization results that improve understanding of radionuclide retention mechanisms, especially for pertechnetate, {sup 99}TcO{sub 4}{sup -} in four different waste forms: Cast Stone, DuraLith alkali aluminosilicate geopolymer, encapsulated fluidized bed steam reforming (FBSR) product, and Ceramicrete phosphate bonded ceramic. These data and results will be used to fill existing data gaps on the candidate technologies to support a decision-making process that will identify a subset of the candidate waste forms that are most promising and should undergo further performance testing.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

  20. FY16 Annual Accomplishments - Waste Form Development and Performance: Evaluation Of Ceramic Waste Forms - Comparison Of Hot Isostatic Pressed And Melt Processed Fabrication Methods

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dandeneau, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-10-13

    FY16 efforts were focused on direct comparison of multi-phase ceramic waste forms produced via melt processing and HIP methods. Based on promising waste form compositions previously devised at SRNL[13], simulant material was prepared at SRNL and a portion was sent to the Australian Nuclear Science and Technology Organization (ANSTO) for HIP treatments, while the remainder of the material was melt processed at SRNL. The microstructure, phase formation, elemental speciation, and leach behavior, and radiation stability of the fabricated ceramics was performed. In addition, melt-processed ceramics designed with different fractions of hollandite, zirconolite, perovskite, and pyrochlore phases were investigated. for performance and properties. Table 1 lists the samples studied.

  1. Standard test method for splitting tensile strength for brittle nuclear waste forms

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1989-01-01

    1.1 This test method is used to measure the static splitting tensile strength of cylindrical specimens of brittle nuclear waste forms. It provides splitting tensile-strength data that can be used to compare the strength of waste forms when tests are done on one size of specimen. 1.2 The test method is applicable to glass, ceramic, and concrete waste forms that are sufficiently homogeneous (Note 1) but not to coated-particle, metal-matrix, bituminous, or plastic waste forms, or concretes with large-scale heterogeneities. Cementitious waste forms with heterogeneities >1 to 2 mm and 5 mm can be tested using this procedure provided the specimen size is increased from the reference size of 12.7 mm diameter by 6 mm length, to 51 mm diameter by 100 mm length, as recommended in Test Method C 496 and Practice C 192. Note 1—Generally, the specimen structural or microstructural heterogeneities must be less than about one-tenth the diameter of the specimen. 1.3 This test method can be used as a quality control chec...

  2. Finite element analysis of ion transport in solid state nuclear waste form materials

    Science.gov (United States)

    Rabbi, F.; Brinkman, K.; Amoroso, J.; Reifsnider, K.

    2017-09-01

    Release of nuclear species from spent fuel ceramic waste form storage depends on the individual constituent properties as well as their internal morphology, heterogeneity and boundary conditions. Predicting the release rate is essential for designing a ceramic waste form, which is capable of effectively storing the spent fuel without contaminating the surrounding environment for a longer period of time. To predict the release rate, in the present work a conformal finite element model is developed based on the Nernst Planck Equation. The equation describes charged species transport through different media by convection, diffusion, or migration. And the transport can be driven by chemical/electrical potentials or velocity fields. The model calculates species flux in the waste form with different diffusion coefficient for each species in each constituent phase. In the work reported, a 2D approach is taken to investigate the contributions of different basic parameters in a waste form design, i.e., volume fraction, phase dispersion, phase surface area variation, phase diffusion co-efficient, boundary concentration etc. The analytical approach with preliminary results is discussed. The method is postulated to be a foundation for conformal analysis based design of heterogeneous waste form materials.

  3. Alternative Electrochemical Salt Waste Forms, Summary of FY11-FY12 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mccloy, John S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lepry, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rodriguez, Carmen P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Windisch, Charles F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westman, Matthew P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rieck, Bennett T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lang, Jesse B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Olszta, Matthew J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pierce, David A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-01-17

    The Fuel Cycle Research and Development Program, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is currently investigating alternative waste forms for wastes generated from nuclear fuel processing. One such waste results from an electrochemical separations process, called the “Echem” process. The Echem process utilizes a molten KCl-LiCl salt to dissolve the fuel. This process results in a spent salt containing alkali, alkaline earth, lanthanide halides and small quantities of actinide halides, where the primary halide is chloride with a minor iodide fraction. Pacific Northwest National Laboratory (PNNL) is concurrently investigating two candidate waste forms for the Echem spent-salt: high-halide minerals (i.e., sodalite and cancrinite) and tellurite (TeO2)-based glasses. Both of these candidates showed promise in fiscal year (FY) 2009 and FY2010 with a simplified nonradioactive simulant of the Echem waste. Further testing was performed on these waste forms in FY2011 and FY2012 to assess the possibility of their use in a sustainable fuel cycle. This report summarizes the combined results from FY2011 and FY2012 efforts.

  4. Alternative Electrochemical Salt Waste Forms, Summary of FY11-FY12 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Mccloy, John S.; Crum, Jarrod V.; Lepry, William C.; Rodriguez, Carmen P.; Windisch, Charles F.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Olszta, Matthew J.; Pierce, David A.

    2014-03-26

    The Fuel Cycle Research and Development Program, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is currently investigating alternative waste forms for wastes generated from nuclear fuel processing. One such waste results from an electrochemical separations process, called the “Echem” process. The Echem process utilizes a molten KCl-LiCl salt to dissolve the fuel. This process results in a spent salt containing alkali, alkaline earth, lanthanide halides and small quantities of actinide halides, where the primary halide is chloride with a minor iodide fraction. Pacific Northwest National Laboratory (PNNL) is concurrently investigating two candidate waste forms for the Echem spent-salt: high-halide minerals (i.e., sodalite and cancrinite) and tellurite (TeO2)-based glasses. Both of these candidates showed promise in fiscal year (FY) 2009 and FY2010 with a simplified nonradioactive simulant of the Echem waste. Further testing was performed on these waste forms in FY2011 and FY2012 to assess the possibility of their use in a sustainable fuel cycle. This report summarizes the combined results from FY2011 and FY2012 efforts.

  5. Secondary Waste Form Screening Test Results—Cast Stone and Alkali Alumino-Silicate Geopolymer

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Eric M.; Cantrell, Kirk J.; Westsik, Joseph H.; Parker, Kent E.; Um, Wooyong; Valenta, Michelle M.; Serne, R. Jeffrey

    2010-06-28

    PNNL is conducting screening tests on the candidate waste forms to provide a basis for comparison and to resolve the formulation and data needs identified in the literature review. This report documents the screening test results on the Cast Stone cementitious waste form and the Geopolymer waste form. Test results suggest that both the Cast Stone and Geopolymer appear to be viable waste forms for the solidification of the secondary liquid wastes to be treated in the ETF. The diffusivity for technetium from the Cast Stone monoliths was in the range of 1.2 × 10-11 to 2.3 × 10-13 cm2/s during the 63 days of testing. The diffusivity for technetium from the Geopolymer was in the range of 1.7 × 10-10 to 3.8 × 10-12 cm2/s through the 63 days of the test. These values compare with a target of 1 × 10-9 cm2/s or less. The Geopolymer continues to show some fabrication issues with the diffusivities ranging from 1.7 × 10-10 to 3.8 × 10-12 cm2/s for the better-performing batch to from 1.2 × 10-9 to 1.8 × 10-11 cm2/s for the poorer-performing batch. In the future more comprehensive and longer term performance testing will be conducted, to further evaluate whether or not these waste forms will meet the regulation and performance criteria needed to cost-effectively dispose of secondary wastes.

  6. Physicochemical properties and morphology of vitreous waste forms incorporating hazardous incineration ash

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Tae; Park, Hyun Soo; Kim, Joon Hyung [KAERI, Taejon (Korea, Republic of); Koo, Ja Kong [KAIST, Taejon (Korea, Republic of); Seo, Yong Chil [Yonsei University, Seoul (Korea, Republic of)

    1999-07-01

    Ash melting experiments were conducted to investigate the applicability of glass matrix as a binder for the solidification of hazardous incineration ash. Several batches of hazardous incineration ash from a paint-factory were melt at 1300 deg C. to fabrication solidified waste forms with the addition of different contents of base-glass material as an additive. The XRD analysis of the final waste forms indicated mixtures of ash and additive are satisfactorily vitrified to form amorphous phases. Even though solidification agents (base-glass) were added, the total waste volume was reduced after vitrification. The volume reduction factor increased with HWI ash loading and reached up t 4.6. The minimum compressive strength and microhardness were 54 MPA and 5.9 GPa, respectively, which were higher than those of cement-solidified incineration ash. All the vitreous waste forms passes the standard extraction tests performed in accordance with Korean MOE's EP and US EPA's TCLP method and thus they could be classified as non-hazardous wastes to save disposal cost. The total mass leach rates were several g/m{sup 2}.d after 14 days of MCC-5S leaching test. Morphology and chemical analysis of waste glass by SEM/EDS before and after leaching tests showed that titanium in the glass network was very durable to leave a Ti-rich layer at the surface of the waste form after leaching. The overall assessment of experimental results showed that the applicability of vitrification technology to treat hazardous incineration ashes would be viable. (author). 20 refs., 6 tabs., 7 figs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  8. Secondary Waste Form Screening Test Results—THOR® Fluidized Bed Steam Reforming Product in a Geopolymer Matrix

    Energy Technology Data Exchange (ETDEWEB)

    Pires, Richard P.; Westsik, Joseph H.; Serne, R. Jeffrey; Mattigod, Shas V.; Golovich, Elizabeth C.; Valenta, Michelle M.; Parker, Kent E.

    2011-07-14

    Screening tests are being conducted to evaluate waste forms for immobilizing secondary liquid wastes from the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Plans are underway to add a stabilization treatment unit to the Effluent Treatment Facility to provide the needed capacity for treating these wastes from WTP. The current baseline is to use a Cast Stone cementitious waste form to solidify the wastes. Through a literature survey, DuraLith alkali-aluminosilicate geopolymer, fluidized-bed steam reformation (FBSR) granular product encapsulated in a geopolymer matrix, and a Ceramicrete phosphate-bonded ceramic were identified both as candidate waste forms and alternatives to the baseline. These waste forms have been shown to meet waste disposal acceptance criteria, including compressive strength and universal treatment standards for Resource Conservation and Recovery Act (RCRA) metals (as measured by the toxicity characteristic leaching procedure [TCLP]). Thus, these non-cementitious waste forms should also be acceptable for land disposal. Information is needed on all four waste forms with respect to their capability to minimize the release of technetium. Technetium is a radionuclide predicted to be in the secondary liquid wastes in small quantities, but the Integrated Disposal Facility (IDF) risk assessment analyses show that technetium, even at low mass, produces the largest contribution to the estimated IDF disposal impacts to groundwater.

  9. Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Weber, William J. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering; Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering

    2016-09-20

    This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effects of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-04-15

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

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

    Science.gov (United States)

    Saleh, H. M.; Tawfik, M. E.; Bayoumi, T. A.

    2011-04-01

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

  13. Glass waste forms for heat-generating Cs{sup +} and Sr{sup 2+} wastes from pyro-processing

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Min Suk; Heo, Jong [POSTECH, Pohang (Korea, Republic of); Park, Hwan Seo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Pyro-processing is one of the promising recycling technologies for spent nuclear fuel (SNF) from Light Water Reactors (LWR) in Korea. This processing is able to separate radioactive waste nuclei and reduce heat loading in storage site by extraction of heat generating radioactive nuclei. In this study, we used alumino-borosilicate glasses for the immobilization of Cs{sub 2}O and SrO wastes. Glasses were prepared and their important properties including chemical durability were analyzed. In addition, heat generation and its effect on thermal stability of glasses was examined. Glass waste forms that contain heat-generating Cs{sup +} and Sr{sup 2+} from pyro-processing were synthesized. Basic properties of glasses such as densities, linear expansion coefficients and glass-transition temperatures were similar to those of industrial radioactive waste glass. Analysis on the heat load simulation under the failure of the cooling system indicated that maximum temperature inside the canisters are well below the glass-transition temperature of each glass.

  14. Radiation stability test on multiphase glass ceramic and crystalline ceramic waste forms

    Science.gov (United States)

    Tang, Ming; Kossoy, Anna; Jarvinen, Gordon; Crum, Jarrod; Turo, Laura; Riley, Brian; Brinkman, Kyle; Fox, Kevin; Amoroso, Jake; Marra, James

    2014-05-01

    A radiation stability study was performed on glass ceramic and crystalline ceramic waste forms. These materials are candidate host materials for immobilizing alkali/alkaline earth (Cs/Sr-CS) + lanthanide (LN) + transition metal (TM) fission product waste streams from nuclear fuel reprocessing. In this study, glass ceramics were fabricated using a borosilicate glass as a matrix in which to incorporate CS/LN/TM combined waste streams. The major phases in these multiphase materials are powellite, oxyaptite, pollucite, celsian, and durable residual glass phases. Al2O3 and TiO2 were combined with these waste components to produce multiphase crystalline ceramics containing hollandite-type phases, perovskites, pyrochlores and other minor metal titanate phases. For the radiation stability test, selected glass ceramic and crystalline ceramic samples were exposed to different irradiation environments including low fluxes of high-energy (∼1-5 MeV) protons and alpha particles generated by an ion accelerator, high fluxes of low-energy (hundreds of keV) krypton particles generated by an ion implanter, and in-situ electron irradiations in a transmission electron microscope. These irradiation experiments were performed to simulate self-radiation effects in a waste form. Ion irradiation-induced microstructural modifications were examined using X-ray diffraction and transmission electron microscopy. Our preliminary results reveal different radiation tolerance in different crystalline phases under various radiation damage environments. However, their stability may be rate dependent which may limit the waste loading that can be achieved.

  15. Chemical durability and degradation mechanisms of HT9 based alloy waste forms with variable Zr content

    Energy Technology Data Exchange (ETDEWEB)

    Olson, L. N. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-30

    In Corrosion studies were undertaken on alloy waste forms that can result from advanced electrometallurgical processing techniques to better classify their durability and degradation mechanisms. The waste forms were based on the RAW3-(URe) composition, consisting primarily of HT9 steel and other elemental additions to simulate nuclear fuel reprocessing byproducts. The solution conditions of the corrosion studies were taken from an electrochemical testing protocol, and meant to simulate conditions in a repository. The alloys durability was examined in alkaline and acidic brines.

  16. Feasibility of metallurgical waste encapsulation in a clay formed matrix

    Science.gov (United States)

    Juhnevica, I.; Kucinska, J.; Sardiko, A.; Mezinskis, G.

    2011-12-01

    As a result of Joint Stock Company "Liepajas Metalurgs" production process there are produced certain quantity of substances that are harmful for environment and have to be encapsulated into stable structures. Company's target is modification of these substances into products that form stable compounds in order to avoid metal release in environment. Geopolymers can be synthesized from many materials with a high concentration of aluminosilicates such as metakaolin or fly ash. Heavy metal immobilization in geopolymeric structures is not thought to be caused by physical encapsulation alone, but also through adsorption of the metal ions into the geopolymer structure and possibly even bonding of the metal ions into the structure. All samples have been analyzed with X-Ray, FTIR spectroscopy; chemical analysis and compressive strength tests have been performed. Chemical analysis of geopolymeric samples shows that the main component leached from samples during the boiling in water is Na2O that can be explained by more alkaline components nature - Na2SiO3, NaOH, and SO3. Fe2O3 and ZnO are not detected in water extracts at all samples.

  17. Summary Report: Glass-Ceramic Waste Forms for Combined Fission Products

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-23

    Glass-ceramic waste form development began in FY 2010 examining two combined waste stream options: (1) alkaline earth (CS) + lanthanide (Ln), and (2) + transition metal (TM) fission-product waste streams generated by the uranium extraction (UREX+) separations process. Glass-ceramics were successfully developed for both options however; Option 2 was selected over Option 1, at the conclusion of 2010, because Option 2 immobilized all three waste streams with only a minimal decrease in waste loading. During the first year, a series of three glass (Option 2) were fabricated that varied waste loading-WL (42, 45, and 50 mass%) at fixed molar ratios of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali both at 1.75. These glass-ceramics were slow cooled and characterized in terms of phase assemblage and preliminary irradiation stability. This fiscal year, further characterization was performed on the FY 2010 Option 2 glass-ceramics in terms of: static leach testing, phase analysis by transmission electron microscopy (TEM), and irradiation stability (electron and ion). Also, a new series of glass-ceramics were developed for Option 2 that varied the additives: Al{sub 2}O{sub 3} (0-6 mass%), molar ratio of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali (1.75 to 2.25) and waste loading (50, 55, and 60 mass%). Lastly, phase pure powellite and oxyapatite were synthesized for irradiation studies. Results of this fiscal year studies showed compositional flexibility, chemical stability, and radiation stability in the current glass-ceramic system. First, the phase assemblages and microstructure of all of the FY 2010 and 2011 glass-ceramics are very similar once subjected to the slow cool heat treatment. The phases identified in these glass-ceramics were oxyapatite, powellite, cerianite, and ln-borosilicate. This shows that variations in waste loading or additives can be accommodated without drastically changing the phase assemblage of the waste form, thus making the processing and performance

  18. Hanford Waste Vitrification Plant Quality Assurance Program description for high-level waste form development and qualification. Revision 3, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    The Hanford Waste Vitrification Plant Project has been established to convert the high-level radioactive waste associated with nuclear defense production at the Hanford Site into a waste form suitable for disposal in a deep geologic repository. The Hanford Waste Vitrification Plant will mix processed radioactive waste with borosilicate material, then heat the mixture to its melting point (vitrification) to forin a glass-like substance that traps the radionuclides in the glass matrix upon cooling. The Hanford Waste Vitrification Plant Quality Assurance Program has been established to support the mission of the Hanford Waste Vitrification Plant. This Quality Assurance Program Description has been written to document the Hanford Waste Vitrification Plant Quality Assurance Program.

  19. DEVELOPMENT OF CERAMIC WASTE FORMS FOR AN ADVANCED NUCLEAR FUEL CYCLE

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-11-30

    A series of ceramic waste forms were developed and characterized for the immobilization of a Cesium/Lanthanide (CS/LN) waste stream anticipated to result from nuclear fuel reprocessing. Simple raw materials, including Al{sub 2}O{sub 3} and TiO{sub 2} were combined with simulated waste components to produce multiphase ceramics containing hollandite-type phases, perovskites (particularly BaTiO{sub 3}), pyrochlores and other minor metal titanate phases. 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. X-Ray Diffraction (XRD) and Scanning Electron Microscopy coupled with Energy Dispersive Spectroscopy (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. Identification of excess Al{sub 2}O{sub 3} via XRD and 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. The Phase II compositions generally contained a reduced amount of unreacted Al{sub 2}O{sub 3} as identified by XRD. 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.

  20. Plutonium-238 alpha-decay damage study of the ceramic waste form.

    Energy Technology Data Exchange (ETDEWEB)

    Frank, S M [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Barber, T L [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Cummings, D G [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; DiSanto, T [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Esh, D W [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; Giglio, J J [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Goff, K M [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Johnson, S G [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Kennedy, J R [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Jue, J-F [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Noy, M [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; O' Holleran, T P [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Sinkler, W [UOP LLC, 25 E Algonquin Road, Des Plaines, IL 60017

    2006-03-27

    An accelerated alpha-decay damage study of a glass-bonded sodalite ceramic waste form has recently been completed. The purpose of this study was to investigate the physical and chemical durability of the waste form after significant exposure to alpha decay. This accelerated alpha-decay study was performed by doping the ceramic waste form with {sup 238}Pu which has a much greater specific activity than {sup 239}Pu that is normally present in the waste form. The alpha-decay dose at the end of the four year study was approximately 1 x 10{sup 18} alpha-decays/gram of material. An equivalent time period for a similar dose of {sup 239}Pu would require approximately 1100 years. After four years of exposure to {sup 238}Pu alpha decay, the investigation observed little change to the physical or chemical durability of the ceramic waste form (CWF). Specifically, the {sup 238}Pu-loaded CWF maintained it's physical integrity, namely that the density remained constant and no cracking or phase de-bonding was observed. The materials chemical durability and phase stability also did not change significantly over the duration of the study. The only significant measured change was an increase of the unit-cell lattice parameters of the plutonium oxide and sodalite phases of the material and an increase in the release of salt components and plutonium of the waste form during leaching tests, but, as mentioned, these did not lead to any overall loss of waste form durability. The principal findings from this study are: (1) {sup 238}Pu-loaded CWF is similar in microstructure and phase composition to referenced waste form. (2) Pu was observed primarily as oxide comprised of aggregates of nano crystals with aggregates ranging in size from submicron to twenty microns in diameter. (3) Pu phases were primarily found in the intergranular glassy regions. (4) PuO phase shows expected unit cell volume expansion due to alpha decay damage of approximately 0.7%, and the sodalite phase unit cell

  1. On the Durability of Nuclear Waste Forms from the Perspective of Long-Term Geologic Repository Performance

    Directory of Open Access Journals (Sweden)

    Yifeng Wang

    2013-12-01

    Full Text Available High solid/water ratios and slow water percolation cause the water in a repository to quickly (on a repository time scale reach radionuclide solubility controlled by the equilibrium with alteration products; the total release of radionuclides then becomes insensitive to the dissolution rates of primary waste forms. It is therefore suggested that future waste form development be focused on conditioning waste forms or repository environments to minimize radionuclide solubility, rather than on marginally improving the durability of primary waste forms.

  2. On-line Technology Information System (OTIS): Solid Waste Management Technology Information Form (SWM TIF)

    Science.gov (United States)

    Levri, Julie A.; Boulanger, Richard; Hogan, John A.; Rodriguez, Luis

    2003-01-01

    Contents include the following: What is OTIS? OTIS use. Proposed implementation method. Development history of the Solid Waste Management (SWM) Technology Information Form (TIF) and OTIS. Current development state of the SWM TIF and OTIS. Data collection approach. Information categories. Critiques/questions/feedback.

  3. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 16. Repository preconceptual design studies: BPNL waste forms in salt

    Energy Technology Data Exchange (ETDEWEB)

    1978-04-01

    This volume, Volume 16, ''Repository Preconceptual Design Studies: BPNL Waste Forms in Salt,'' is one of a 23 volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provide a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This document describes a preconceptual design for a nuclear waste storage facility in salt. The waste forms assumed to arrive at the repository were supplied by Battelle Pacific Northwest Laboratories (BPNL). The facility design consists of several chambers excavated deep within a geologic formation together with access shafts and supportive surface structures. The facility design provides for: receiving and unloading waste containers; lowering them down shafts to the mine level; transporting them to the proper storage area and emplacing them in mined storage rooms. Drawings of the facility design are contained in TM-36/17, ''Drawings for Repository Preconceptual Design Studies: BPNL Waste Forms in Salt.''

  4. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 16. Repository preconceptual design studies: BPNL waste forms in salt

    Energy Technology Data Exchange (ETDEWEB)

    1978-04-01

    This volume, Volume 16, ''Repository Preconceptual Design Studies: BPNL Waste Forms in Salt,'' is one of a 23 volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provide a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This document describes a preconceptual design for a nuclear waste storage facility in salt. The waste forms assumed to arrive at the repository were supplied by Battelle Pacific Northwest Laboratories (BPNL). The facility design consists of several chambers excavated deep within a geologic formation together with access shafts and supportive surface structures. The facility design provides for: receiving and unloading waste containers; lowering them down shafts to the mine level; transporting them to the proper storage area and emplacing them in mined storage rooms. Drawings of the facility design are contained in TM-36/17, ''Drawings for Repository Preconceptual Design Studies: BPNL Waste Forms in Salt.''

  5. SCALE UP OF CERAMIC WASTE FORMS FOR THE EBR-II SPENT FUEL TREATMENT PROCESS

    Energy Technology Data Exchange (ETDEWEB)

    Matthew C. Morrison; Kenneth J. Bateman; Michael F. Simpson

    2010-11-01

    ABSTRACT SCALE UP OF CERAMIC WASTE FORMS FOR THE EBR-II SPENT FUEL TREATMENT PROCESS Matthew C. Morrison, Kenneth J. Bateman, Michael F. Simpson Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 The ceramic waste process is the intended method for disposing of waste salt electrolyte, which contains fission products from the fuel-processing electrorefiners (ER) at the INL. When mixed and processed with other materials, the waste salt can be stored in a durable ceramic waste form (CWF). The development of the CWF has recently progressed from small-scale testing and characterization to full-scale implementation and experimentation using surrogate materials in lieu of the ER electrolyte. Two full-scale (378 kg and 383 kg) CWF test runs have been successfully completed with final densities of 2.2 g/cm3 and 2.1 g/cm3, respectively. The purpose of the first CWF was to establish material preparation parameters. The emphasis of the second pre-qualification test run was to evaluate a preliminary multi-section CWF container design. Other considerations were to finalize material preparation parameters, measure the material height as it consolidates in the furnace, and identify when cracking occurs during the CWF cooldown process.

  6. Development of a Waste Treatment Process to Deactivate Reactive Uranium Metal and Produce a Stable Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Gates-Anderson, D D; Laue, C A; Fitch, T E

    2002-01-17

    This paper highlights the results of initial investigations conducted to support the development of an integrated treatment process to convert pyrophoric metallic uranium wastes to a non-pyrophoric waste that is acceptable for land disposal. Several dissolution systems were evaluated to determine their suitability to dissolve uranium metal and that yield a final waste form containing uranium specie(s) amenable to precipitation, stabilization, adsorption, or ion exchange. During initial studies, one gram aliquots of uranium metal or the uranium alloy U-2%Mo were treated with 5 to 60 mL of selected reagents. Treatment systems screened included acids, acid mixtures, and bases with and without addition of oxidants. Reagents used included hydrochloric, sulfuric, nitric, and phosphoric acids, sodium hypochlorite, sodium hydroxide and hydrogen peroxide. Complete dissolution of the uranium turnings was achieved with the H{sub 3}PO{sub 4}/HCI system at room temperature within minutes. The sodium hydroxide/hydrogen peroxide, and sodium hypochlorite systems achieved complete dissolution but required elevated temperatures and longer reaction times. A ranking system based on criteria, such as corrosiveness, temperature, dissolution time, off-gas type and amount, and liquid to solid ratio, was designed to determine the treatment systems that should be developed further for a full-scale process. The highest-ranking systems, nitric acid/sulfuric acid and hydrochloric acid/phosphoric acid, were given priority in our follow-on investigations.

  7. Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Eric M.; McGrail, B. Peter; Rodriguez, Elsa A.; Schaef, Herbert T.; Saripalli, Prasad; Serne, R. Jeffrey; Krupka, Kenneth M.; Martin, P. F.; Baum, Steven R.; Geiszler, Keith N.; Reed, Lunde R.; Shaw, Wendy J.

    2004-09-01

    This data package documents the experimentally derived input data on the representative waste glasses; LAWA44, LAWB45, and LAWC22. This data will be used for Subsurface Transport Over Reactive Multi-phases (STORM) simulations of the Integrated Disposal Facility (IDF) for immobilized low-activity waste (ILAW). The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in July 2005. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali (Na+)-hydrogen (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 (PUF) and product consistency (PCT) tests where used for accelerated weathering or aging of the glasses in order to determine a chemical reaction network of secondary phases that form. The majority of the thermodynamic data used in this data package were extracted from the thermody-namic database package shipped with the geochemical code EQ3/6, version 8.0. Because of the expected importance of 129I release from secondary waste streams being sent to IDF from various thermal treatment processes, parameter estimates for diffusional release and solubility-controlled release from cementitious waste forms were estimated from the available literature.

  8. Yucca Mountain project canister material corrosion studies as applied to the electrometallurgical treatment metallic waste form

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D.

    1996-11-01

    Yucca Mountain, Nevada is currently being evaluated as a potential site for a geologic repository. As part of the repository assessment activities, candidate materials are being tested for possible use as construction materials for waste package containers. A large portion of this testing effort is focused on determining the long range corrosion properties, in a Yucca Mountain environment, for those materials being considered. Along similar lines, Argonne National Laboratory is testing a metallic alloy waste form that also is scheduled for disposal in a geologic repository, like Yucca Mountain. Due to the fact that Argonne`s waste form will require performance testing for an environment similar to what Yucca Mountain canister materials will require, this report was constructed to focus on the types of tests that have been conducted on candidate Yucca Mountain canister materials along with some of the results from these tests. Additionally, this report will discuss testing of Argonne`s metal waste form in light of the Yucca Mountain activities.

  9. Stability of ceramic waste forms in potential repository environments: a review

    Energy Technology Data Exchange (ETDEWEB)

    Johnston, R. J.; Palmer, R. A.

    1982-03-31

    Most scenarios for geologic disposal of high-level nuclear waste include the eventual intrusion of groundwater into the repository. Reactions in the system and eventual release of the radionuclides, if any, will be controlled by the chemistry of the groundwater, the surrounding rock, the waste form, and any engineered barrier materials that are present, as well as by the temperature and pressure of the system. This report is a compilation and evaluation of the work completed to date on interactions within the waste-form/host-rock/groundwater system at various points in its lifetime. General results from leaching experiments are presented as a basis for comparison. The factors involved in studying the complete system are discussed so that future research may avoid some of the oversights of past research. Although relatively little hard data on prototype waste-form/repository-system interactions exist at this time, the available data and their implications are discussed. Sorption studies and models for predicting radionuclide migration are also presented, again with a study of the factors involved.

  10. Glass composite waste forms for iodine confined in bismuth-embedded SBA-15

    Science.gov (United States)

    Yang, Jae Hwan; Park, Hwan Seo; Ahn, Do-Hee; Yim, Man-Sung

    2016-11-01

    The aim of this study was to stabilize bismuth-embedded SBA-15 that captured iodine gas by fabrication of monolithic waste forms. The iodine containing waste was mixed with Bi2O3 (a stabilizing additive) and low-temperature sintering glass followed by pelletizing and the sintering process to produce glass composite materials. Iodine volatility during the sintering process was significantly affected by the ratio of Bi2O3 and the glass composition. It was confirmed that BiI3, the main iodine phase within bismuth-embedded SBA-15, was effectively transformed to the mixed phases of Bi5O7I and BiOI. The initial leaching rates of iodine from the glass composite waste forms ranged 10-3-10-2 g/m2 day, showing the stability of the iodine phases encapsulated by the glassy networks. It was also observed that common groundwater anions (e.g., chloride, carbonate, sulfite, and fluoride) elevated the iodine leaching rate by anion exchange reactions. The present results suggest that the glass composite waste form of bismuth-embedded SBA-15 could be a candidate material for stable storage of 129I.

  11. Yucca Mountain project canister material corrosion studies as applied to the electrometallurgical treatment metallic waste form

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D.

    1996-11-01

    Yucca Mountain, Nevada is currently being evaluated as a potential site for a geologic repository. As part of the repository assessment activities, candidate materials are being tested for possible use as construction materials for waste package containers. A large portion of this testing effort is focused on determining the long range corrosion properties, in a Yucca Mountain environment, for those materials being considered. Along similar lines, Argonne National Laboratory is testing a metallic alloy waste form that also is scheduled for disposal in a geologic repository, like Yucca Mountain. Due to the fact that Argonne`s waste form will require performance testing for an environment similar to what Yucca Mountain canister materials will require, this report was constructed to focus on the types of tests that have been conducted on candidate Yucca Mountain canister materials along with some of the results from these tests. Additionally, this report will discuss testing of Argonne`s metal waste form in light of the Yucca Mountain activities.

  12. A development of the stabilization technology for the solid form of radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. G.; Lee, Y. H.; Lee, K. M.; Ann, S. J.; Son, J. S. [KAERI, Taejon (Korea, Republic of)

    2003-07-01

    In this study, a modified bituminization technology has been developed, which needs no grinding of the granular resin waste, and enables the solid form to keep its shape stability as good as that of a cemented solid form. Also, the study intended to apply the developed technology to the practical treatment of radioactive resin waste. In the experiment, the granular type resin was used and the straight-run distillation bitumen with penetration rate 60/70 was used as the solidifying agent. The PE was used as the additive. The shape stability increased remarkably with the additive of PE, which act as a binder in the solid form. The shape of the solid from was maintained without failure during the long-term exposure test when the additive content of spent PE is more than 10 wt %. The proper ranges of bitumen content, PE content and operating temperature are 30-50 wt %, 10-20 wt % and 180.deg.C respectively. The bituminized solid form of radioactive resin waste by the technology of this study has the remarkably superior quality than the conventional solid forms, partially for the shape stability.

  13. PRELIMINARY ASSESSMENT OF THE LOW-TEMPERATURE WASTE FORM TECHNOLOGY COUPLED WITH TECHNETIUM REMOVAL

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.

    2014-05-13

    The U.S. Department of Energy Office of Environmental Management (EM) is engaging the national laboratories to provide the scientific and technological rigor to support EM program and project planning, technology development and deployment, project execution, and assessment of program outcomes. As an early demonstration of this new responsibility, Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) have been chartered to implement a science and technology program addressing low-temperature waste forms for immobilization of DOE aqueous waste streams, including technetium removal as an implementing technology. As a first step, the laboratories examined the technical risks and uncertainties associated with the Cast Stone waste immobilization projects at Hanford. Science and technology needs were identified for work associated with 1) conducting performance assessments and risk assessments of waste form and disposal system performance, and 2) technetium chemistry in tank wastes and separations of technetium from waste processing streams. Technical approaches to address the science and technology needs were identified and an initial sequencing priority was suggested. The following table summarizes the most significant science and technology needs and associated approaches to address those needs. These approaches and priorities will be further refined and developed as strong integrated teams of researchers from national laboratories, contractors, industry, and academia are brought together to provide the best science and technology solutions. Implementation of a science and technology program that addresses these needs by pursuing the identified approaches will have immediate benefits to DOE in reducing risks and uncertainties associated with near-term decisions regarding supplemental immobilization at Hanford. Longer term, the work has the potential for cost savings and for providing a strong technical foundation for future

  14. A simplified method for estimation of jarosite and acid-forming sulfates in acid mine wastes.

    Science.gov (United States)

    Li, Jun; Smart, Roger St C; Schumann, Russell C; Gerson, Andrea R; Levay, George

    2007-02-01

    In acid base accounting (ABA) estimates of acid mine wastes, the acid potential (AP) estimate can be improved by using the net carbonate value (NCV) reactive sulfide S method rather than total S assay methods but this does not give recovery of potentially acid producing ferrous and ferric sulfates present in many wastes. For more accurate estimation of AP, an effective, site-specific method to quantify acid sulfate salts, such as jarosite and melanterite, in waste rocks has been developed and tested on synthetic and real wastes. The SPOCAS (acid sulfate soils) methods have been modified to an effective, rapid method to speciate sulfate forms in different synthetic waste samples. A three-step sequential extraction procedure has been established. These steps are: (1) argon-purged water extraction (3 min) to extract soluble Fe(II) salts (particularly melanterite), epsomite and gypsum (1 wt.% S) as copper sulfides, the second step of roasting needs to be excluded from the procedure with an increased time of 4 M HCl extraction to 16 h for jarosite determination.

  15. Waste Form and Indrift Colloids-Associated Radionuclide Concentrations: Abstraction and Summary

    Energy Technology Data Exchange (ETDEWEB)

    R. Aguilar

    2003-06-24

    This Model Report describes the analysis and abstractions of the colloids process model for the waste form and engineered barrier system components of the total system performance assessment calculations to be performed with the Total System Performance Assessment-License Application model. Included in this report is a description of (1) the types and concentrations of colloids that could be generated in the waste package from degradation of waste forms and the corrosion of the waste package materials, (2) types and concentrations of colloids produced from the steel components of the repository and their potential role in radionuclide transport, and (3) types and concentrations of colloids present in natural waters in the vicinity of Yucca Mountain. Additionally, attachment/detachment characteristics and mechanisms of colloids anticipated in the repository are addressed and discussed. The abstraction of the process model is intended to capture the most important characteristics of radionuclide-colloid behavior for use in predicting the potential impact of colloid-facilitated radionuclide transport on repository performance.

  16. Corrosion behavior of technetium waste forms exposed to various aqueous environments

    Energy Technology Data Exchange (ETDEWEB)

    Kolman, David Gary [Los Alamos National Laboratory; Jarvinen, Gordon [Los Alamos National Laboratory; Mausolf, Edward [UNIV OF NEVADA; Czerwinski, Ken [UNIV OF NEVADA; Poineau, Frederic [UNIV OF NEVADA

    2009-01-01

    Technetium is a long-lived beta emitter produced in high yields from uranium as a waste product in spent nuclear fuel and has a high degree of environmental mobility as pertechnetate. It has been proposed that Tc be immobilized into various metallic waste forms to prevent Tc mobility while producing a material that can withstand corrosion exposed to various aqueous medias to prevent the leachability of Tc to the environment over long periods of time. This study investigates the corrosion behavior of Tc and Tc alloyed with 316 stainless steel and Zr exposed to a variety of aqueous media. To date, there is little investigative work related to Tc corrosion behavior and less related to potential Tc containing waste forms. Results indicate that immobilizing Tc into stainless steel-zirconium alloys can be a promising technique to store Tc for long periods of time while reducing the need to separately store used nuclear fuel cladding. Initial results indicate that metallic Tc and its alloys actively corrode in all media. We present preliminary corrosion rates of 100% Tc, 10% Tc - 90% SS{sub 85%}Zr{sub 15%}, and 2% Tc - 98% SS{sub 85%}Zr{sub 15%} in varying concentrations of nitric acid and pH 10 NaOH using the resistance polarization method while observing the trend that higher concentrations of Tc alloyed to the sample tested lowers the corrosion rate of the proposed waste package.

  17. INNOVATIVE TECHNIQUES AND TECHNOLOGY APPLICATION IN MANAGEMENT OF REMOTE HANDLED AND LARGE SIZED MIXED WASTE FORMS

    Energy Technology Data Exchange (ETDEWEB)

    BLACKFORD LT

    2008-02-04

    of RCRA storage regulations, reduce costs for waste management by nearly 50 percent, and create a viable method for final treatment and disposal of these waste forms that does not impact retrieval project schedules. This paper is intended to provide information to the nuclear and environmental clean-up industry with the experience of CH2M HILL and ORP in managing these highly difficult waste streams, as well as providing an opportunity for sharing lessons learned, including technical methods and processes that may be applied at other DOE sites.

  18. Secondary Waste Form Down-Selection Data Package—DuraLith

    Energy Technology Data Exchange (ETDEWEB)

    Mattigod, Shas V.; Westsik, Joseph H.

    2011-09-15

    This data package developed for the DuraLith wasteform includes information available in the open literature and from data obtained from testing currently underway. DuraLith is an alkali-activated geopolymer waste form developed by the Vitreous State Laboratory at The Catholic University of America (VSL-CUA) for encapsulating liquid radioactive waste. A DuraLith waste form developed for treating Hanford secondary waste liquids is prepared by alkali-activation of a mixture of ground blast furnace slag and metakaolinite with sand used as a filler material. Based on optimization tests, solid waste loading of {approx}7.5% and {approx}14.7 % has been achieved using the Hanford secondary waste S1 and S4 simulants, respectively. The Na loading in both cases is equivalent to {approx}6 M. Some of the critical parameters for the DuraLith process include, hydrogen generation and heat evolution during activator solution preparation using the waste simulant, heat evolution during and after mixing the activator solution with the dry ingredients, and a working window of {approx}20 minutes to complete the pouring of the DuraLith mixture into molds. Results of the most recent testing indicated that the working window can be extended to {approx}30 minutes if 75 wt% of the binder components, namely, blast furnace slag and metakaolin are replaced by Class F fly ash. A preliminary DuraLith process flow sheet developed by VSL-CUA for processing Hanford secondary waste indicated that 10 to 22 waste monoliths (each 48 ft3 in volume) can be produced per day. There are no current pilot-scale or full-scale DuraLith plants under construction or in operation; therefore, the cost of DuraLith production is unknown. The results of the non-regulatory leach tests, EPA Draft 1313 and 1316, Waste Simulant S1-optimized DuraLith specimens indicated that the concentrations of RCRA metals (Ag, Cd, Cr, Hg, and Pb) in the leachates were well below the Universal Treatment Standard limits in 40 CFR 268

  19. Immobilization of noble metal fission products in a metallic waste form

    Energy Technology Data Exchange (ETDEWEB)

    Frank, S. M.; Bateman, K.; Marsden, K. C.; Keiser, D. D.; O' Holleran, T. P.; Hahn, P. A. [Idaho National Laboratory, Boise (United States)

    2008-08-15

    Development of the metallic waste form for the consolidation of spent-fuel cladding and the immobilization of specific fission-product radionuclides occurred as part of the larger Electrometallurgical Treatment Research and Demonstration conducted by Argonne National Laboratory for the U.S. Department of Energy from 1996 to 1999. The Global Nuclear Energy Partnership (GNEP) proposal for advanced reprocessing of spent nuclear fuel also proposes to combine recovered fission-product technetium and other transition metal fission products, primarily the undissolved solid (UDS) residue from the dissolver vessels, into a metallic, high-level waste form for geological disposal. This approach is similar to the production of the MWF produced during the treatment of spent EBR-II fuel at the INL.

  20. Cold crucible induction melter studies for making glass ceramic waste forms: A feasibility assessment

    Science.gov (United States)

    Crum, Jarrod; Maio, Vince; McCloy, John; Scott, Clark; Riley, Brian; Benefiel, Brad; Vienna, John; Archibald, Kip; Rodriguez, Carmen; Rutledge, Veronica; Zhu, Zihua; Ryan, Joe; Olszta, Matthew

    2014-01-01

    Glass ceramics are being developed to immobilize fission products, separated from used nuclear fuel by aqueous reprocessing, into a stable waste form suitable for disposal in a geological repository. This work documents the glass ceramic formulation at bench scale and for a scaled melter test performed in a pilot-scale (∼1/4 scale) cold crucible induction melter (CCIM). Melt viscosity, electrical conductivity, and crystallization behavior upon cooling were measured on a small set of compositions to select a formulation for melter testing. Property measurements also identified a temperature range for melter operation and cooling profiles necessary to crystallize the targeted phases in the waste form. Bench scale and melter run results successfully demonstrate the processability of the glass ceramic using the CCIM melter technology.

  1. Separations and Waste Forms Research and Development FY 2013 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-12-01

    The Separations and Waste Form Campaign (SWFC) under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Program (FCRD) is responsible for developing advanced separation and waste form technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year (FY) 2013 accomplishments report provides a highlight of the results of the research and development (R&D) efforts performed within SWFC in FY 2013. Each section contains a high-level overview of the activities, results, technical point of contact, applicable references, and documents produced during the fiscal year. This report briefly outlines campaign management and integration activities, but the intent of the report is to highlight the many technical accomplishments made during FY 2013.

  2. Direct Measurement of Surface Dissolution Rates in Potential Nuclear Waste Forms: The Example of Pyrochlore.

    Science.gov (United States)

    Fischer, Cornelius; Finkeldei, Sarah; Brandt, Felix; Bosbach, Dirk; Luttge, Andreas

    2015-08-19

    The long-term stability of ceramic materials that are considered as potential nuclear waste forms is governed by heterogeneous surface reactivity. Thus, instead of a mean rate, the identification of one or more dominant contributors to the overall dissolution rate is the key to predict the stability of waste forms quantitatively. Direct surface measurements by vertical scanning interferometry (VSI) and their analysis via material flux maps and resulting dissolution rate spectra provide data about dominant rate contributors and their variability over time. Using pyrochlore (Nd2Zr2O7) pellet dissolution under acidic conditions as an example, we demonstrate the identification and quantification of dissolution rate contributors, based on VSI data and rate spectrum analysis. Heterogeneous surface alteration of pyrochlore varies by a factor of about 5 and additional material loss by chemo-mechanical grain pull-out within the uppermost grain layer. We identified four different rate contributors that are responsible for the observed dissolution rate range of single grains. Our new concept offers the opportunity to increase our mechanistic understanding and to predict quantitatively the alteration of ceramic waste forms.

  3. EVALUATION OF THOR MINERALIZED WASTE FORMS FOR THE DOE ADVANCED REMEDIATION TECHNOLOGIES PHASE 2 PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C.; Jantzen, C.

    2012-02-02

    The U.S. Department of Energy's (DOE) 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. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). 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 Cs-137, I-129, Tc-99, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750 C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates

  4. EVALUATION OF THOR MINERALIZED WASTE FORMS FOR THE DOE ADVANCED REMEDIATION TECHNOLOGIES PHASE 2 PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C.; Jantzen, C.

    2012-02-02

    The U.S. Department of Energy's (DOE) 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. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). 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 Cs-137, I-129, Tc-99, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750 C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  6. USING CENTER HOLE HEAT TRANSFER TO REDUCE FORMATION TIMES FOR CERAMIC WASTE FORMS FROM PYROPROCESSING

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth J. Bateman; Charles W. Solbrig

    2006-07-01

    The waste produced from processing spent fuel from the EBR II reactor must be processed into a waste form suitable for long term storage in Yucca Mountain. The method chosen produces zeolite granules mixed with glass frit, which must then be converted into a solid. This is accomplished by loading it into a can and heating to 900 C in a furnace regulated at 915 C. During heatup to 900 C, the zeolite and glass frit react and consolidate to produce a sodalite monolith. The resultant ceramic waste form (CWF) is then cooled. The waste is 52 cm in diameter and initially 300 cm long but consolidates to 150 cm long during the heating process. After cooling it is then inserted in a 5-DHLW/DOE SNF Long Canister. Without intervention, the waste takes 82 hours to heat up to 900 C in a furnace designed to geometrically fit the cylindrical waste form. This paper investigates the reduction in heating times possible with four different methods of additional heating through a center hole. The hole size is kept small to maximize the amount of CWF that is processed in a single run. A hole radius of 1.82 cm was selected which removes only 1% of the CWF. A reference computation was done with a specified inner hole surface temperature of 915 C to provide a benchmark for the amount of improvement which can be made. It showed that the heatup time can potentially be reduced to 43 hours with center hole heating. The first method, simply pouring high temperature liquid aluminum into the hole, did not produce any noticeable effect on reducing heat up times. The second method, flowing liquid aluminum through the hole, works well as long as the velocity is high enough (2.5 cm/sec) to prevent solidification of the aluminum during the initial front movement of the aluminum into the center hole. The velocity can be reduced to 1 cm/sec after the initial front has traversed the ceramic. This procedure reduces the formation time to near that of the reference case. The third method, flowing a gas

  7. A science-based approach to understanding waste form durability in open and closed nuclear fuel cycles

    Science.gov (United States)

    Peters, M. T.; Ewing, R. C.

    2007-05-01

    There are two compelling reasons for understanding source term and near-field processes in a radioactive waste geologic repository. First, almost all of the radioactivity is initially in the waste form, mainly in the spent nuclear fuel (SNF) or nuclear waste glass. Second, over long periods, after the engineered barriers are degraded, the waste form is a primary control on the release of radioactivity. Thus, it is essential to know the physical and chemical state of the waste form after hundreds of thousands of years. The United States Department of Energy's Yucca Mountain Repository Program has initiated a long-term program to develop a basic understanding of the fundamental mechanisms of radionuclide release and a quantification of the release as repository conditions evolve over time. Specifically, the research program addresses four critical areas: (a) SNF dissolution mechanisms and rates; (b) formation and properties of U6+-secondary phases; (c) waste form-waste package interactions in the near-field; and (d) integration of in-package chemical and physical processes. The ultimate goal is to integrate the scientific results into a larger scale model of source term and near-field processes. This integrated model will be used to provide a basis for understanding the behaviour of the source term over long time periods (greater than 105 years). Such a fundamental and integrated experimental and modelling approach to source term processes can also be readily applied to development of advanced waste forms as part of a closed nuclear fuel cycle. Specifically, a fundamental understanding of candidate waste form materials stability in high temperature/high radiation environments and near-field geochemical/hydrologic processes could enable development of advanced waste forms 'tailored' to specific geologic settings.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-11-01

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

  9. Cold crucible induction melter test for crystalline ceramic waste form fabrication: A feasibility assessment

    Science.gov (United States)

    Amoroso, Jake W.; Marra, James; Dandeneau, Christopher S.; Brinkman, Kyle; Xu, Yun; Tang, Ming; Maio, Vince; Webb, Samuel M.; Chiu, Wilson K. S.

    2017-04-01

    The first scaled proof-of-principle cold crucible induction melter (CCIM) test to process a multiphase ceramic waste form from a simulated combined (Cs/Sr, lanthanide and transition metal fission products) commercial used nuclear fuel waste stream was recently conducted in the United States. X-ray diffraction, 2-D X-ray absorption near edge structure (XANES), electron microscopy, inductively coupled plasma-atomic emission spectroscopy (and inductively coupled plasma-mass spectroscopy for Cs), and product consistency tests were used to characterize the fabricated CCIM material. Characterization analyses confirmed that a crystalline ceramic with a desirable phase assemblage was produced from a melt using a CCIM. Primary hollandite, pyrochlore/zirconolite, and perovskite phases were identified in addition to minor phases rich in Fe, Al, or Cs. The material produced in the CCIM was chemically homogeneous and displayed a uniform phase assemblage with acceptable aqueous chemical durability.

  10. STABILIZING GLASS BONDED WASTE FORMS CONTAINING FISSION PRODUCTS SEPARATED FROM SPENT NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth J. Bateman; Charles W. Solbrig

    2008-07-01

    A model has been developed to represent the stresses developed when a molten, glass-bonded brittle cylinder (used to store nuclear material) is cooled from high temperature to working temperature. Large diameter solid cylinders are formed by heating glass or glass-bonded mixtures (mixed with nuclear waste) to high temperature (915°C). These cylinders must be cooled as the final step in preparing them for storage. Fast cooling time is desirable for production; however, if cooling is too fast, the cylinder can crack into many pieces. To demonstrate the capability of the model, cooling rate cracking data were obtained on small diameter (7.8 cm diameter) glass-only cylinders. The model and experimental data were combined to determine the critical cooling rate which separates the non-cracking stable glass region from the cracked, non-stable glass regime. Although the data have been obtained so far only on small glass-only cylinders, the data and model were used to extrapolate the critical-cooling rates for large diameter ceramic waste form (CWF) cylinders. The extrapolation estimates long term cooling requirements. While a 52-cm diameter cylinder (EBR-II-waste size) can be cooled to 100°C in 70 hours without cracking, a 181.5-cm diameter cylinder (LWR waste size) requires 35 days to cool to 100°C. These cooling times are long enough that verification of these estimates are required so additional experiments are planned on both glass only and CWF material.

  11. Preliminary evaluation of alternative waste form solidification processes. Volume I. Identification of the processes.

    Energy Technology Data Exchange (ETDEWEB)

    Treat, R.L.; Nesbitt, J.F.; Blair, H.T.; Carter, J.G.; Gorton, P.S.; Partain, W.L.; Timmerman, C.L.

    1980-04-01

    This document contains preconceptual design data on 11 processes for the solidification and isolation of nuclear high-level liquid wastes (HLLW). The processes are: in-can glass melting (ICGM) process, joule-heated glass melting (JHGM) process, glass-ceramic (GC) process, marbles-in-lead (MIL) matrix process, supercalcine pellets-in-metal (SCPIM) matrix process, pyrolytic-carbon coated pellets-in-metal (PCCPIM) matrix process, supercalcine hot-isostatic-pressing (SCHIP) process, SYNROC hot-isostatic-pressing (SYNROC HIP) process, titanate process, concrete process, and cermet process. For the purposes of this study, it was assumed that each of the solidification processes is capable of handling similar amounts of HLLW generated in a production-sized fuel reprocessing plant. It was also assumed that each of the processes would be enclosed in a shielded canyon or cells within a waste facility located at the fuel reprocessing plant. Finally, it was assumed that all of the processes would be subject to the same set of regulations, codes and standards. Each of the solidification processes converts waste into forms that may be acceptable for geological disposal. Each process begins with the receipt of HLLW from the fuel reprocessing plant. In this study, it was assumed that the original composition of the HLLW would be the same for each process. The process ends when the different waste forms are enclosed in canisters or containers that are acceptable for interim storage. Overviews of each of the 11 processes and the bases used for their identification are presented in the first part of this report. Each process, including its equipment and its requirements, is covered in more detail in Appendices A through K. Pertinent information on the current state of the art and the research and development required for the implementation of each process are also noted in the appendices.

  12. Cordierite containing ceramic membranes from smectetic clay using natural organic wastes as pore-forming agents

    Directory of Open Access Journals (Sweden)

    W. Misrar

    2017-06-01

    Full Text Available Cordierite ceramic membranes were manufactured from natural clay, oxides and organic wastes as pore forming agents. Mixtures aforementioned materials with the pore-forming agents (up to 10 wt.% were investigated in the range 1000–1200 °C using thermal analysis, X-ray diffraction, scanning electron microscopy, mercury porosimetry and filtration tests. Physical properties (density, water absorption and bending strength were correlated to the processing factors (pore-forming agent addition, firing temperature and soaking time. The results showed that cordierite together with spinel, diopside and clinoenstatite neoformed. SEM analysis revealed heterogeneous aspects. The results of the response surface methodology showed that the variations of physical properties versus processing parameters were well described by the used polynomial model. The addition of pore forming agent and temperature were the most influential factors. Filtration tests were performed on the best performing sample. The results allowed to testify that these membranes could be used in waste water treatment.

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

    Energy Technology Data Exchange (ETDEWEB)

    Singh, D.; Wagh, A.; Knox, L. [Argonne National Lab., Argonne, IL (United States); Mayberry, J. [Science Applications International Corp., Idaho Falls, ID (United States)

    1994-03-01

    Chemically bonded phosphate ceramics are proposed as candidates for solidification and stabilization of some of the {open_quotes}problem{close_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.

  14. Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model

    Energy Technology Data Exchange (ETDEWEB)

    Denia Djokic; Steven J. Piet; Layne F. Pincock; Nick R. Soelberg

    2013-02-01

    This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system , and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity.

  15. Studies of high-level waste form performance at Japan Atomic Energy Research Institute

    Energy Technology Data Exchange (ETDEWEB)

    Banba, Tsunetaka; Mitamura, Hisayoshi; Kuramoto, Kenichi; Kamizono, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Inagaki, Yahohiro

    1998-02-01

    The JAERI studies on the properties of the glass and ceramic waste forms, which have been done in the last several years, are described briefly. For the long-term evaluation of glass waste form performance under repository condition, leachability has studied from the standpoints of understanding of alteration layers, effects of groundwater and effects of redox condition using the radioactive or non-radioactive glass samples. The studies revealed that (1) the reactions in the alteration layers, such as crystal growth, continue after the apparent release of elements from the glass almost ceases, (2) under somewhat reducing conditions, Fe dissolves easily into leachates, and hydrated silicate surface layer tends to dissolve more easily with Fe in reduced synthetic groundwater than in deionized water, (3) precipitation of PuO{sub 2}{center_dot}xH{sub 2}O(am) is controlling the leaching of soluble species of Pu under both redox conditions, and the dominant soluble species is Pu(OH){sub 4}{sup 0} under reducing condition. Ceramics are considered as most promising materials for the actinide-rich wastes arising from partitioning and transmutation processes because of their outstanding durability for long term. In the present study, {alpha}-decay damage effects on the density and leaching behavior of perovskite (1 of 3 main minerals forming Synroc) were investigated by an accelerated experiment using the actinide doping technique. A decrease in density of Cm-doped perovskite reaches 1.3% at a dose of 9x10{sup 17} {alpha}-decays{center_dot}g{sup -1}. The leach rate of perovskite increases with an increase in accumulated {alpha}-decay doses. Application of zirconia- and alumina-based ceramics for incorporating actinides was also investigated by inactive laboratory tests with an emphasis on crystallographic phase stability and chemical durability. The yttria-stabilized zirconia is stable crystallographically in the wide ranges of Ce and/or Nd content and have excellent

  16. Waste Form Release Calculations for the 2005 Integrated Disposal Facility Performance Assessment. Erratum

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-06

    This report refers to or contains Kg values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected Kg values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

  17. Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment Erratum

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-06

    This report refers to or contains Kg values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected Kg values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

  18. Waste Form Release Calculations for the 2005 Integrated Disposal Facility Performance Assessment Erratum

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-06

    This report refers to or contains Kg values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected Kg values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

  19. Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. Erratum

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-06

    This report refers to or contains Kg values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected Kg values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

  20. Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Weber, William [Univ. of Tennessee, Knoxville, TN (United States)

    2016-09-20

    This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effects of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied. Two model used-fuel materials, nanograined CeO2 and ZrO2, were fabricated as part of this study. To complement the work on damage evolution in nanocrystalline CeO2 and ZrO2 during helium implantation and heavy ion irradiation, additional irradiations were performed on single crystal CeO2 and ZrO2. Samples were irradiated to ion fluences corresponding to an irradiation dose ranging from 0.11 to 100 dpa (displacements per atom), which is comparable to the irradiated dose expected during interim and long-term storage. Detailed transmission electron microscopy, Rutherford backscattering and Raman spectroscopy analysis have been carried out on these irradiated materials. The critical helium concentration for formation of helium bubbles was found to be 0.15 atomic percent (at%) in these samples, which is similar to that found in 238Pu-doped UO2. This critical helium concentration for bubble formation will be achieved in less than 100 years for MOX used fuels, in about 1000

  1. Separation of tc from Uranium and development of metallic Technetium waste forms

    Science.gov (United States)

    Mausolf, Edward John

    The isotope Technetium-99 (99Tc) is a major fission product of the nuclear industry. In the last decade, approximately 20 tons of 99Tc have been produced by the US nuclear industry. Due to its long half-life (t1/2 = 214,000 yr), beta radiotoxicity, and high mobility as pertechnetate [TcO4]-, Tc represents long-term concern to the biosphere. Various options have been considered to manage 99Tc. One of them is its separation from spent fuel, conversion to the metal and incorporation into a metallic waste form for long-term disposal. After dissolution of spent fuel in nitric acid and extraction of U and Tc in organic media, previously developed methods can be used to separate Tc from U, convert the separate Tc stream to the metal and reuse the uranium component of the fuel. A variety of metallic waste forms, ranging from pure Tc metal to ternary Tc alloys combined with stainless steel (SS) and Zr are proposed. The goal of this work was to examine three major questions: What is the optimal method to separate Tc from U? After separation, what is the most efficient method to convert the Tc stream to Tc metal? Finally, what is the corrosion behavior of Tc metal, Tc-SS alloys and Tc-Zr-SS alloys in 0.01M NaCl? The goal is to predict the long term behavior of Tc metallic waste in a hypothetical storage environment. In this work, three methods have been used to separate Tc from U: anionic exchange resin, liquid-liquid extraction and precipitation. Of the three methods studied, anionic exchange resins is the most selective. After separation of Tc from U, three different methods were studied to convert the Tc stream to the metal: thermal treatment under hydrogen atmosphere, electrochemical and chemical reduction of pertechnetate in aqueous media. The thermal treatment of the Tc stream under hydrogen atmosphere is the preferred method to produce Tc metal. After Tc metal is isolated, it will be incorporated into a metal host phase. Three different waste forms were produced for

  2. Microstructural analysis and corrosion behavior of zirconium-stainless steel metallic waste form

    Energy Technology Data Exchange (ETDEWEB)

    Das, N., E-mail: nirupamd@barc.gov.in; Abraham, G.; Sengupta, P.; Arya, Ashok; Kain, V.; Dey, G.K.

    2015-12-15

    Management of radioactive metallic waste using “alloy melting route” is currently being investigated by several researchers. In the present study, potentiodynamic polarizations were conducted on six as-cast zirconium (Zr)-stainless steel (SS) alloys (i.e. Zr-25, 20, 16, 12, 8 and 5 wt.% SS) at pH = 1, 5 and 8. Electrochemical behavior of metallic-waste-form (MWF) alloys containing more than 16 wt.% SS showed lower potentials at the break down of passivity attributed to localized attack mainly at Cr-depleted matrix–intermetallic interfaces. Zr–5SS and Zr–12SS alloys contain Zr{sub 3}(Fe, Cr, Ni)/Zr{sub 3}(Fe, Cr)-type of phases and their interfaces with matrices were prone to localized attack. Whereas, Zr–8SS and Zr–16SS alloys demonstrated better corrosion resistance in comparison to Zr–5SS and Zr–12SS respectively. In addition, occurrence of Laves phase, e.g. Zr{sub 2}(Fe, Cr), in Zr–8SS and Zr–16SS alloys makes them suitable for MWF. - Highlights: • Acceptable SS content in Zr–SS metallic waste form alloy is limited to 16 wt.%. • Localized attack was observed at the Cr-depleted intermetallics–matrix interfaces. • Zr-8 wt.% SS showed highest corrosion resistance among all the Zr–SS alloys. • Zr-16 wt.% SS having sufficient Laves intermetallic phase is preferable for MWF alloy.

  3. Terahertz Time-Domain Spectroscopy for In Situ Monitoring of Ceramic Nuclear Waste Forms

    Science.gov (United States)

    Clark, Braeden M.; Sundaram, S. K.

    2016-10-01

    The use of terahertz time-domain spectroscopy (THz-TDS) is presented as a non-contact method for in situ monitoring of ceramic waste forms. Single-phase materials of zirconolite (CaZrTi2O7), pyrochlore (Nd2Ti2O7), and hollandite (BaCs0.3Cr2.3Ti5.7O16 and BaCs0.3CrFeAl0.3Ti5.7O16) were characterized. The refractive index and dielectric properties in THz frequencies demonstrate the ability to distinguish between these materials. Differences in processing methods show distinct changes in both the THz-TDS spectra and optical and dielectric properties of these ceramic phases. The temperature dependence of the refractive index and relative permittivity of pyrochlore and zirconolite materials in the range of 25-200 °C is found to follow an exponential increasing trend. This can also be used to monitor the temperature of the ceramic waste forms on storage over extended geological time scales.

  4. Fundamental thermodynamics of actinide-bearing mineral waste forms. 1998 annual progress report

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, M.A. [Los Alamos National Lab., NM (US); Ebbinghaus, B.B.

    1998-06-01

    'The end of the Cold War raised the need for the technical community to be concerned with the disposition of excess nuclear weapon material. The plutonium will either be converted into mixed-oxide fuel for use in nuclear reactors or immobilized in glass or ceramic waste forms and placed in a repository. The stability and behavior of plutonium in the ceramic materials as well as the phase behavior and stability of the ceramic material in the environment is not well established. In order to provide technically sound solutions to these issues, thermodynamic data are essential in developing an understanding of the chemistry and phase equilibria of the actinide-bearing mineral waste form materials proposed as immobilization matrices. Mineral materials of interest include zircon, zirconolite, and pyrochlore. High temperature solution calorimetry is one of the most powerful techniques, sometimes the only technique, for providing the fundamental thermodynamic data needed to establish optimum material fabrication parameters, and more importantly, understand and predict the behavior of the mineral materials in the environment. The purpose of this project is to experimentally determine the enthalpy of formation of actinide orthosilicates, the enthalpy of formation of actinide substituted zircon, zirconolite and pyrochlore, and develop an understanding of the bonding characteristics and stability of these materials. This report summarizes work after eight months of a three year project.'

  5. Fundamental Thermodynamics of Actinide-Bearing Mineral Waste Forms - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Mark A.; Ebbinghaus, Bartley B.; Navrotsky, Alexandra

    2001-03-01

    The end of the Cold War raised the need for the technical community to be concerned with the disposition of excess nuclear weapon material. The plutonium will either be converted into mixed-oxide fuel for use in nuclear reactors or immobilized in glass or ceramic waste forms and placed in a repository. The stability and behavior of plutonium in the ceramic materials as well as the phase behavior and stability of the ceramic material in the environment is not well established. In order to provide technically sound solutions to these issues, thermodynamic data are essential in developing an understanding of the chemistry and phase equilibria of the actinide-bearing mineral waste form materials proposed as immobilization matrices. Mineral materials of interest include zircon, zirconolite, and pyrochlore. High temperature solution calorimetry is one of the most powerful techniques, sometimes the only technique, for providing the fundamental thermodynamic data needed to establish optimum material fabrication parameters, and more importantly understand and predict the behavior of the mineral materials in the environment. The purpose of this project is to experimentally determine the enthalpy of formation of actinide orthosilicates, the enthalpies of formation of actinide substituted zirconolite and pyrochlore, and develop an understanding of the bonding characteristics and stabilities of these materials.

  6. Ni and Cr addition to alloy waste forms to reduce radionuclide environmental releases

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-11

    Reference alloy waste forms (RAW) were fabricated and underwent hybrid corrosion/immersion testing to parameterize the ANL analytical oxidative-dissolution model to enable the calculation of fractional release rates and to determine the effectiveness of Ni and Cr trim additions in reducing release rates of radionuclide surrogates. Figure 1 shows the prototypical multiphase microstructure of the alloys with each phase type contributing about equally to the exposed surface area. The waste forms tested at SRNL were variations of the RAW-6 formulation that uses HT9 as the main alloy component, and are meant to enable evaluation of the impact of Ni and Cr trim additions on the release rates of actinides and Tc-99. The test solutions were deaerated alkaline and acidic brines, ranging in pH 3 to pH 10, representing potential repositories with those conditions. The testing approach consisted of 4 major steps; 1) bare surface corrosion measurements at pH values of 3, 5, 8, and 10, 2) hybrid potentiostatic hold/exposure measurements at pH 3, 3) measurement of radionuclide concentrations and relations to anodic current from potentiostatic holds, and 4) identification of corroding phases using SEM/EDS of electrodes.

  7. Extended Development Work to Validate a HLW Calcine Waste Form via INL's Cold Crucible Induction Melter

    Energy Technology Data Exchange (ETDEWEB)

    James A. King; Vince Maio

    2011-09-01

    To accomplish calcine treatment objectives, the Idaho Clean-up Project contractor, CWI, has chosen to immobilize the calcine in a glass-ceramic via the use of a Hot-Isostatic-Press (HIP); a treatment selection formally documented in a 2010 Record of Decision (ROD). Even though the HIP process may prove suitable for the calcine as specified in the ROD and validated in a number of past value engineering sessions, DOE is evaluating back-up treatment methods for the calcine as a result of the technical, schedule, and cost risk associated with the HIPing process. Consequently DOE HQ has requested DOE ID to make INL's bench-scale cold-crucible induction melter (CCIM) available for investigating its viability as a process alternate to calcine treatment. The waste form is the key component of immobilization of radioactive waste. Providing a solid, stable, and durable material that can be easily be stored is the rationale for immobilization of radioactive waste material in glass, ceramic, or glass-ceramics. Ceramic waste forms offer an alternative to traditional borosilicate glass waste forms. Ceramics can usually accommodate higher waste loadings than borosilicate glass, leading to smaller intermediate and long-term storage facilities. Many ceramic phases are known to possess superior chemical durability as compared to borosilicate glass. However, ceramics are generally multiphase systems containing many minor phase that make characterization and prediction of performance within a repository challenging. Additionally, the technologies employed in ceramic manufacture are typically more complex and expensive. Thus, many have proposed using glass-ceramics as compromise between in the more inexpensive, easier to characterize glass waste forms and the more durable ceramic waste forms. Glass-ceramics have several advantages over traditional borosilicate glasses as a waste form. Borosilicate glasses can inadvertently devitrify, leading to a less durable product that could

  8. Distribution and Solubility of Radionuclides and Neutron Absorbers in Waste Forms for Disposition of Plutonium Ash and Scraps, Excess Plutonium, and Miscellaneous Spent Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Denis M. Strachan; Dr. David K. Shuh; Dr. Rodney C. Ewing; Dr. Eric R. Vance

    2002-09-23

    The initial goal of this project was to investigate the solubility of radionuclides in glass and other potential waste forms for the purpose of increasing the waste loading in glass and ceramic waste forms. About one year into the project, the project decided to focus on two potential waste forms - glass at PNNL and itianate ceramics at the Australian Nuclear Science and Technology Organisation (ANSTO).

  9. Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank SX-105 And AN-103) By Fluidized Bed Steam Reformation

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, Carol; Herman, Connie; Crawford, Charles; Bannochie, Christopher; Burket, Paul; Daniel, Gene; Cozzi, Alex; Nash, Charles; Miller, Donald; Missimer, David

    2014-01-10

    One of the immobilization technologies under consideration as a Supplemental Treatment for Hanford’s Low Activity Waste (LAW) is Fluidized Bed Steam Reforming (FBSR). The FBSR technology forms a mineral waste form at moderate processing temperatures thus retaining and atomically bonding the halides, sulfates, and technetium in the mineral phases (nepheline, sodalite, nosean, carnegieite). Additions of kaolin clay are used instead of glass formers and the minerals formed by the FBSR technology offers (1) atomic bonding of the radionuclides and constituents of concern (COC) comparable to glass, (2) short and long term durability comparable to glass, (3) disposal volumes comparable to glass, and (4) higher Na2O and SO{sub 4} waste loadings than glass. The higher FBSR Na{sub 2}O and SO{sub 4} waste loadings contribute to the low disposal volumes but also provide for more rapid processing of the LAW. Recent FBSR processing and testing of Hanford radioactive LAW (Tank SX-105 and AN-103) waste is reported and compared to previous radioactive and non-radioactive LAW processing and testing.

  10. INTERNATIONAL PROGRAM: SUMMARY REPORT ON THE PROPERTIES OF CEMENTITIOUS WASTE FORMS

    Energy Technology Data Exchange (ETDEWEB)

    Harbour, J

    2007-03-02

    This report provides a summary of the results on the properties of cementitious waste forms obtained as part of the International Program. In particular, this report focuses on the results of Task 4 of the Program that was initially entitled ''Improved Retention of Key Contaminants of Concern in Low Temperature Immobilized Waste Forms''. Task 4 was a joint program between Khlopin Radium Institute and the Savannah River National Laboratory. The task evolved during this period into a study of cementitious waste forms with an expanded scope that included heat of hydration and fate and transport modeling. This report provides the results for Task 4 of the International Program as of the end of FY06 at which time funding for Task 4 was discontinued due to the needs of higher priority tasks within the International Program. Consequently, some of the subtasks were only partially completed, but it was considered important to capture the results up to this point in time. Therefore, this report serves as the closeout report for Task 4. The degree of immobilization of Tc-99 within the Saltstone waste form was measured through monolithic and crushed grout leaching tests. An effective diffusion coefficient of 4.8 x 10{sup -12} (Leach Index of 11.4) was measured using the ANSI/ANS-16.1 protocol which is comparable with values obtained for tank closure grouts using a dilute salt solution. The leaching results show that, in the presence of concentrated salt solutions such as those that will be processed at the Saltstone Production Facility, blast furnace slag can effectively reduce pertechnetate to the immobile +4 oxidation state. Leaching tests were also initiated to determine the degree of immobilization of selenium in the Saltstone waste form. Results were obtained for the upper bound of projected selenium concentration ({approx}5 x 10{sup -3} M) in the salt solution that will be treated at Saltstone. The ANSI/ANS 16.1 leaching tests provided a value for the

  11. Round-robin testing of a reference glass for low-activity waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, W. L.; Wolf, S. F.

    1999-12-06

    A round robin test program was conducted with a glass that was developed for use as a standard test material for acceptance testing of low-activity waste glasses made with Hanford tank wastes. The glass is referred to as the low-activity test reference material (LRM). The program was conducted to measure the interlaboratory reproducibility of composition analysis and durability test results. Participants were allowed to select the methods used to analyze the glass composition. The durability tests closely followed the Product Consistency Test (PCT) Method A, except that tests were conducted at both 40 and 90 C and that parallel tests with a reference glass were not required. Samples of LRM glass that had been crushed, sieved, and washed to remove fines were provided to participants for tests and analyses. The reproducibility of both the composition and PCT results compare favorably with the results of interlaboratory studies conducted with other glasses. From the perspective of reproducibility of analysis results, this glass is acceptable for use as a composition standard for nonradioactive components of low-activity waste forms present at >0.1 elemental mass % and as a test standard for PCTS at 40 and 90 C. For PCT with LRM glass, the expected test results at the 95% confidence level are as follows: (1) at 40 C: pH = 9.86 {+-} 0.96; [B] = 2.30 {+-} 1.25 mg/L; [Na] = 19.7 {+-} 7.3 mg/L; [Si] = 13.7 {+-} 4.2 mg/L; and (2) at 90 C: pH = 10.92 {+-} 0.43; [B] = 26.7 {+-} 7.2 mg/L; [Na] = 160 {+-} 13 mg/L; [Si] = 82.0 {+-} 12.7 mg/L. These ranges can be used to evaluate the accuracy of PCTS conducted at other laboratories.

  12. Preliminary evaluation of alternative waste form solidification processes. Volume II. Evaluation of the processes

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    This Volume II presents engineering feasibility evaluations of the eleven processes for solidification of nuclear high-level liquid wastes (HHLW) described in Volume I of this report. Each evaluation was based in a systematic assessment of the process in respect to six principal evaluation criteria: complexity of process; state of development; safety; process requirements; development work required; and facility requirements. The principal criteria were further subdivided into a total of 22 subcriteria, each of which was assigned a weight. Each process was then assigned a figure of merit, on a scale of 1 to 10, for each of the subcriteria. A total rating was obtained for each process by summing the products of the subcriteria ratings and the subcriteria weights. The evaluations were based on the process descriptions presented in Volume I of this report, supplemented by information obtained from the literature, including publications by the originators of the various processes. Waste form properties were, in general, not evaluated. This document describes the approach which was taken, the developent and application of the rating criteria and subcriteria, and the evaluation results. A series of appendices set forth summary descriptions of the processes and the ratings, together with the complete numerical ratings assigned; two appendices present further technical details on the rating process.

  13. FORM AND AGING OF PLUTONIUM IN SAVANNAH RIVER SITE WASTE TANK 18

    Energy Technology Data Exchange (ETDEWEB)

    Hobbs, D.

    2012-02-24

    This report provides a summary of the effects of aging on and the expected forms of plutonium in Tank 18 waste residues. The findings are based on available information on the operational history of Tank 18, reported analytical results for samples taken from Tank 18, and the available scientific literature for plutonium under alkaline conditions. These findings should apply in general to residues in other waste tanks. However, the operational history of other waste tanks should be evaluated for specific conditions and unique operations (e.g., acid cleaning with oxalic acid) that could alter the form of plutonium in heel residues. Based on the operational history of other tanks, characterization of samples from the heel residues in those tanks would be appropriate to confirm the form of plutonium. During the operational period and continuing with the residual heel removal periods, Pu(IV) is the dominant oxidation state of the plutonium. Small fractions of Pu(V) and Pu(VI) could be present as the result of the presence of water and the result of reactions with oxygen in air and products from the radiolysis of water. However, the presence of Pu(V) would be transitory as it is not stable at the dilute alkaline conditions that currently exists in Tank 18. Most of the plutonium that enters Savannah River Site (SRS) high-level waste (HLW) tanks is freshly precipitated as amorphous plutonium hydroxide, Pu(OH){sub 4(am)} or hydrous plutonium oxide, PuO{sub 2(am,hyd)} and coprecipitated within a mixture of hydrous metal oxide phases containing metals such as iron, aluminum, manganese and uranium. The coprecipitated plutonium would include Pu{sup 4+} that has been substituted for other metal ions in crystal lattice sites, Pu{sup 4+} occluded within hydrous metal oxide particles and Pu{sup 4+} adsorbed onto the surface of hydrous metal oxide particles. The adsorbed plutonium could include both inner sphere coordination and outer sphere coordination of the plutonium. PuO{sub 2

  14. FORM AND AGING OF PLUTONIUM IN SAVANNAH RIVER SITE WASTE TANK 18

    Energy Technology Data Exchange (ETDEWEB)

    Hobbs, D.

    2012-02-24

    This report provides a summary of the effects of aging on and the expected forms of plutonium in Tank 18 waste residues. The findings are based on available information on the operational history of Tank 18, reported analytical results for samples taken from Tank 18, and the available scientific literature for plutonium under alkaline conditions. These findings should apply in general to residues in other waste tanks. However, the operational history of other waste tanks should be evaluated for specific conditions and unique operations (e.g., acid cleaning with oxalic acid) that could alter the form of plutonium in heel residues. Based on the operational history of other tanks, characterization of samples from the heel residues in those tanks would be appropriate to confirm the form of plutonium. During the operational period and continuing with the residual heel removal periods, Pu(IV) is the dominant oxidation state of the plutonium. Small fractions of Pu(V) and Pu(VI) could be present as the result of the presence of water and the result of reactions with oxygen in air and products from the radiolysis of water. However, the presence of Pu(V) would be transitory as it is not stable at the dilute alkaline conditions that currently exists in Tank 18. Most of the plutonium that enters Savannah River Site (SRS) high-level waste (HLW) tanks is freshly precipitated as amorphous plutonium hydroxide, Pu(OH){sub 4(am)} or hydrous plutonium oxide, PuO{sub 2(am,hyd)} and coprecipitated within a mixture of hydrous metal oxide phases containing metals such as iron, aluminum, manganese and uranium. The coprecipitated plutonium would include Pu{sup 4+} that has been substituted for other metal ions in crystal lattice sites, Pu{sup 4+} occluded within hydrous metal oxide particles and Pu{sup 4+} adsorbed onto the surface of hydrous metal oxide particles. The adsorbed plutonium could include both inner sphere coordination and outer sphere coordination of the plutonium. PuO{sub 2

  15. Initial Evaluation of Processing Methods for an Epsilon Metal Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Strachan, Denis M.; Zumhoff, Mac R.

    2012-06-11

    During irradiation of nuclear fuel in a reactor, the five metals, Mo, Pd, Rh, Ru, and Tc, migrate to the fuel grain boundaries and form small metal particles of an alloy known as epsilon metal ({var_epsilon}-metal). When the fuel is dissolved in a reprocessing plant, these metal particles remain behind with a residue - the undissolved solids (UDS). Some of these same metals that comprise this alloy that have not formed the alloy are dissolved into the aqueous stream. These metals limit the waste loading for a borosilicate glass that is being developed for the reprocessing wastes. Epsilon metal is being developed as a waste form for the noble metals from a number of waste streams in the aqueous reprocessing of used nuclear fuel (UNF) - (1) the {var_epsilon}-metal from the UDS, (2) soluble Tc (ion-exchanged), and (3) soluble noble metals (TRUEX raffinate). Separate immobilization of these metals has benefits other than allowing an increase in the glass waste loading. These materials are quite resistant to dissolution (corrosion) as evidenced by the fact that they survive the chemically aggressive conditions in the fuel dissolver. Remnants of {var_epsilon}-metal particles have survived in the geologically natural reactors found in Gabon, Africa, indicating that they have sufficient durability to survive for {approx} 2.5 billion years in a reducing geologic environment. Additionally, the {var_epsilon}-metal can be made without additives and incorporate sufficient foreign material (oxides) that are also present in the UDS. Although {var_epsilon}-metal is found in fuel and Gabon as small particles ({approx}10 {micro}m in diameter) and has survived intact, an ideal waste form is one in which the surface area is minimized. Therefore, the main effort in developing {var_epsilon}-metal as a waste form is to develop a process to consolidate the particles into a monolith. Individually, these metals have high melting points (2617 C for Mo to 1552 C for Pd) and the alloy is

  16. Effect Of Oxidation On Chromium Leaching And Redox Capacity Of Slag-Containing Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-03-01

    The rate of oxidation is important to the long-term performance of reducing salt waste forms because the solubility of some contaminants, e.g., technetium, is a function of oxidation state. TcO4- in the salt solution is reduced to Tc(IV) and has been shown to react with ingredients in the waste form to precipitate low solubility sulfide and/or oxide phases [Shuh, et al., 1994, Shuh, et al., 2000, Shuh, et al., 2003]. Upon exposure to oxygen, the compounds containing Tc(IV) oxidize to the pertechnetate ion, Tc(VII)O4-, which is very soluble. Consequently the rate of technetium oxidation front advancement into a monolith and the technetium leaching profile as a function of depth from an exposed surface are important to waste form performance and ground water concentration predictions. An approach for measuring contaminant oxidation rate (effective contaminant specific oxidation rate) based on leaching of select contaminants of concern is described in this report. In addition, the relationship between reduction capacity and contaminant oxidation is addressed. Chromate was used as a non-radioactive surrogate for pertechnetate in simulated waste form samples. Depth discrete subsamples were cut from material exposed to Savannah River Site (SRS) field cured conditions. The subsamples were prepared and analyzed for both reduction capacity and chromium leachability. Results from field-cured samples indicate that the depth at which leachable chromium was detected advanced further into the sample exposed for 302 days compared to the sample exposed to air for 118 days (at least 50 mm compared to at least 20 mm). Data for only two exposure time intervals is currently available. Data for additional exposure times are required to develop an equation for the oxidation front progression. Reduction capacity measurements (per the Angus-Glasser method, which is a measurement of the ability of a material to chemically reduce Ce(IV) to Ce

  17. Implications of transmutation on the defect chemistry in crystalline waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Uberuaga, B.P., E-mail: blas@lanl.go [Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Jiang, C.; Stanek, C.R.; Sickafus, K.E. [Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Marks, N.A. [Nanochemistry Research Institute, Curtin University of Technology, P.O. Box U1987, Perth, WA 6845 (Australia); Carter, D.J.; Rohl, A.L. [Nanochemistry Research Institute, Curtin University of Technology, P.O. Box U1987, Perth, WA 6845 (Australia); iVEC, Technology Park, Kensington, WA 6151 (Australia)

    2010-10-01

    Radioactive decay within the solid state creates chemical environments which are typically incommensurate with the initial host structure. Using a combined theoretical and computational approach, we discuss this 'transmutation problem' in the context of the short-lived fission products Cs-137 and Sr-90. We show how a Kroeger-Vink treatment is insufficient for understanding defects arising from transmutation, and present density functional theory data for chemical evolution within two prototypical hosts, CsCl and SrTiO{sub 3}. While the latter has a strong driving force for phase separation with increasing Zr content, the Cs(Ba)Cl system is surprisingly stable. The sharp difference between these two findings points to the need for better understanding of novel chemistry in nuclear waste forms.

  18. AN INITIAL ASSESSMENT OF POTENTIAL PRODUCTION TECHNOLOGIES FOR EPSILON-METAL WASTE FORMS

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, Aashish; Strachan, Denis M.

    2011-03-01

    This report examines and ranks a total of seven materials processing techniques that may be potentially utilized to consolidate the undissolved solids from nuclear fuel reprocessing into a low-surface area form. Commercial vendors of processing equipment were contacted and literature researched to gather information for this report. Typical equipment and their operation, corresponding to each of the seven techniques, are described in the report based upon the discussions and information provided by the vendors. Although the report does not purport to describe all the capabilities and issues of various consolidation techniques, it is anticipated that this report will serve as a guide by highlighting the key advantages and disadvantages of these techniques. The processing techniques described in this report were broadly classified into those that employed melting and solidification, and those in which the consolidation takes place in the solid-state. Four additional techniques were examined that were deemed impractical, but were included for completeness. The techniques were ranked based on criteria such as flexibility in accepting wide-variety of feed-stock (chemistry, form, and quantity), ease of long-term maintenance, hot cell space requirements, generation of additional waste streams, cost, and any special considerations. Based on the assumption of ~2.5 L of waste to be consolidated per day, sintering based techniques, namely, microwave sintering, spark plasma sintering and hot isostatic pressing, were ranked as the top-3 choices, respectively. Melting and solidification based techniques were ranked lower on account of generation of volatile phases and difficulties associated with reactivity and containment of the molten metal.

  19. NRC nuclear waste management technical support in the development of nuclear waste form criteria. Task 4. Test development review

    Energy Technology Data Exchange (ETDEWEB)

    Czyscinski, K.S.; Swyler, K.J.; Klamut, C.J.

    1980-05-01

    This interim report concerns the development of testing procedures to assess the performance of waste packages to be used for high-level waste disposal in geologic repositories. Single component testing of the waste package is determined to be a workable strategy for testing and evaluation in terms of NRC release rate criteria. An initial literature review has identified key tests and those variables which must be included in testing procedures to simulate repository conditions. The range of these conditions remains to be determined precisely. Methods for leach, corrosion, and sorption testing are reviewed and initial recommendations made for preferred procedures. A combination of static and dynamic tests is needed to evaluate waste package component performance. Additional research is necessary in certain areas both to establish reliable testing methods and to define the range of testing variables. Research recommendations are included in the report. Ancillary measurements will be required to ensure that key tests rigorously assess the durability of waste package components under anticipated repository conditions. In particular, radiation effects in the repository environment must be considered and, where necessary, simulated during critical testing. Research is recommended to aid in determining when and how this should be done.

  20. Optimized data flow for the waste form documentation of compactable radioactive wastes; Optimierter Datenfluss zur Erstellung von Abfallgebindedokumentationen fuer pressbare radioaktive Abfaelle

    Energy Technology Data Exchange (ETDEWEB)

    Lange, M. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Risch, R. [WTI Wissenschaftlich-Technische Ingenieurleistung GmbH, Juelich (Germany)

    2010-05-15

    According to the national radiation protection regulations (Strahlenschutzverordnung)it is necessary to perform a detailed documentation on radioactive materials, including the waste flow, transport and storage. The book-keeping system AVK is an authorized data base system. The authors describe the data relevant sheets, the activity calculation procedure based on local dose rate measurements and gamma spectroscopy and the necessary documents for the licensing procedure. For a structured and efficient waste form documentation a data base (PIKA-AS: project information and control system) was developed by GNS that includes all documentation relevant data from different measuring and calculation activities.

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

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H.; Serne, R. Jeffrey; Pierce, Eric M.; Cozzi, Alex; Chung, Chul-Woo; Swanberg, David J.

    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.

  2. Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank Farm Blend) By Fluidized Bed Steam Reformation (FBSR)

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Bannochie, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Burket, P. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hall, H. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Missimer, D. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2013-08-01

    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. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Hanford Tank 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. Supplemental Treatment is likely to be required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP’s LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750°C) continuous method by which LAW can be processed irrespective of whether the waste contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be comparable to LAW glass, i.e. leaches Tc-99, Re and Na at <2g/m2 during ASTM C1285 (Product Consistency) durability testing. Monolithing of the granular FBSR product was investigated to prevent dispersion during transport or burial/storage. Monolithing in an inorganic geopolymer binder, which is

  3. INTERNATIONAL PROGRAM: SUMMARY REPORT ON THE PROPERTIES OF CEMENTITIOUS WASTE FORMS

    Energy Technology Data Exchange (ETDEWEB)

    Harbour, J

    2007-03-02

    This report provides a summary of the results on the properties of cementitious waste forms obtained as part of the International Program. In particular, this report focuses on the results of Task 4 of the Program that was initially entitled ''Improved Retention of Key Contaminants of Concern in Low Temperature Immobilized Waste Forms''. Task 4 was a joint program between Khlopin Radium Institute and the Savannah River National Laboratory. The task evolved during this period into a study of cementitious waste forms with an expanded scope that included heat of hydration and fate and transport modeling. This report provides the results for Task 4 of the International Program as of the end of FY06 at which time funding for Task 4 was discontinued due to the needs of higher priority tasks within the International Program. Consequently, some of the subtasks were only partially completed, but it was considered important to capture the results up to this point in time. Therefore, this report serves as the closeout report for Task 4. The degree of immobilization of Tc-99 within the Saltstone waste form was measured through monolithic and crushed grout leaching tests. An effective diffusion coefficient of 4.8 x 10{sup -12} (Leach Index of 11.4) was measured using the ANSI/ANS-16.1 protocol which is comparable with values obtained for tank closure grouts using a dilute salt solution. The leaching results show that, in the presence of concentrated salt solutions such as those that will be processed at the Saltstone Production Facility, blast furnace slag can effectively reduce pertechnetate to the immobile +4 oxidation state. Leaching tests were also initiated to determine the degree of immobilization of selenium in the Saltstone waste form. Results were obtained for the upper bound of projected selenium concentration ({approx}5 x 10{sup -3} M) in the salt solution that will be treated at Saltstone. The ANSI/ANS 16.1 leaching tests provided a value for the

  4. 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

  5. Evidence of Technetium and Iodine from a Sodalite-Bearing Ceramic Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Neeway, James J.; Qafoku, Nikolla; Williams, Benjamin D.; Snyder, Michelle MV; Brown, Christopher F.; Pierce, Eric M.

    2016-03-01

    Current plans for nuclear waste vitrification at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) lack the capacity to treat all of the low activity waste (LAW) that is not encapsulated in the vitrified product. Several technologies are being considered to treat the excess LAW. One such technology is Fluidized Bed Steam Reforming (FBSR). The FBSR process results in a granular product composed of feldspathoid mineral phases that immobilize the major components in the LAW as well as other contaminants of concern (COCs), with Tc and I expected to be present in sodalite cages formed during the process. In order to meet compressive strength requirements at the Hanford Integrated Disposal Facility (IDF), the granular product may be encapsulated in a monolith. To demonstrate the ability of the technology to serve the mission of managing excess LAW, Single Pass Flow-Through (SPFT) tests have been performed on non-radioactive granular materials and granular materials encapsulated in a geopolymer binder produced at the engineering- and bench-scale as well as a granular product produced at the bench scale with actual Hanford tank waste. SPFT tests were conducted at 40 °C for durations up to 2 months with a flow-through solution buffered at pH 9. The forward reaction rate of the non-radioactive mineral product dissolution based on Si release for the granular product was measured to be (6.2 ± 2.1) × 10-4 g/m2d for the engineering-scale product and (13 ± 4.9) × 10-4 g/m2d for the bench-scale product. The resulting non-radioactive monoliths showed forward reaction rates based on Si release of (3.4 ± 1.1) × 10-4 g/m2d for the engineering-scale material and (4.2 ± 1.5) × 10-4 g/m2d for the bench-scale material demonstrating that encapsulation of the FBSR granular product in a monolith does not significantly alter the performance of the material. Finally, an FBSR granular product created at the bench scale using actual Hanford LAW gave similar release values

  6. Instructions and Form for Hazardous Waste Generators, Transporters and Treatment, Storage and Disposal Facilities to Obtain an EPA Identification Number (EPA Form 8700-12/Site Identification Form)

    Science.gov (United States)

    This booklet is designed to help you determine if you are subject to requirements under the Resource Conservation and Recovery Act (RCRA) for notifying the U.S. Environmental Protection Agency (EPA) of your regulated waste activities.

  7. Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group.

    Energy Technology Data Exchange (ETDEWEB)

    Collins, E.; Duguid, J.; Henry, R.; Karell, E.; Laidler, J.; McDeavitt, S.; Thompson, M.; Toth, M.; Williamson, M.; Willit, J.

    1999-08-12

    In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD&D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years.

  8. Identification of lead chemical form in mine waste materials by X-ray absorption spectroscopy

    Science.gov (United States)

    Taga, Raijeli L.; Zheng, Jiajia; Huynh, Trang; Ng, Jack; Harris, Hugh H.; Noller, Barry

    2010-06-01

    X-ray absorption spectroscopy (XAS) provides a direct means for measuring lead chemical forms in complex samples. In this study, XAS was used to identify the presence of plumbojarosite (PbFe6(SO4)4(OH)12) by lead L3-edge XANES spectra in mine waste from a small gold mining operation in Fiji. The presence of plumbojarosite in tailings was confirmed by XRD but XANES gave better resolution. The potential for human uptake of Pb from tailings was measured using a physiologically based extract test (PBET), an in-vitro bioaccessibility (BAc) method. The BAc of Pb was 55%. Particle size distribution of tailings indicated that 40% of PM10 particulates exist which could be a potential risk for respiratory effects via the inhalation route. Food items collected in the proximity of the mine site had lead concentrations which exceed food standard guidelines. Lead within the mining lease exceeded sediment guidelines. The results from this study are used to investigate exposure pathways via ingestion and inhalation for potential risk exposure pathways of Pb in that locality. The highest Pb concentration in soil and tailings was 25,839 mg/kg, exceeding the Australian National Environment Protection Measure (NEPM) soil health investigation levels.

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

    Science.gov (United States)

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

    2016-11-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2003-06-12

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

  11. Forms of avoidance and care of waste as subject of legal regulation

    Directory of Open Access Journals (Sweden)

    Šogorov Stevan

    2011-01-01

    Full Text Available Identification of subject of legal regulation of waste as part of environmental protection law is main goal of this article. Author's starting position is that creation of waste is necessary side product of process of humanization of nature and he points out most important methods for solving that problem. Hierarchy of priorities of solving problem of waste is considered as important. First priority is avoidance of creation of waste, second priority is its material and energetic use, and finally its disposal. Relevant provisions of Waste managing Act of Republic of Serbia of 2009 are argument for acceptance of that hierarchy. Yet there are possible and acceptable exceptions regarding application of existing hierarchy and they represent final part of this article.

  12. Remaining Sites Verification Package for the 100-D-2 Lead Sheeting Waste Site, Waste Site Reclassification Form 2007-030

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-03-19

    The 100-D-2 Lead Sheeting waste site was located approximately 50 m southwest of the 185-D Building and approximately 16 m north of the east/west oriented road. The site consisted of a lead sheet covering a concrete pad. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  13. Dissolution models for glassy waste forms; Modeles de dissolution des formes de dechets a base de verre

    Energy Technology Data Exchange (ETDEWEB)

    Harvey, K.B.

    1995-06-01

    As a part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), a suite of models has been developed that describe the dissolution of a glass under a variety of conditions. This work had two aims: to develop and present the models in such a way that the equations associated with models could be used to unambiguously extract the fundamental dissolution constants of a glass from experimental data, and to demonstrate the correspondence between models and experiments over a sufficiently broad range of conditions such that the models could be used with confidence to forecast performance under conditions that might not be realistically accessible to experiments.

  14. Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank Farm Blend) By Fluidized Bed Steam Reformation (FBSR)

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Bannochie, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Burket, P. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hall, H. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Missimer, D. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2013-08-01

    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. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Hanford Tank 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. Supplemental Treatment is likely to be required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP’s LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750°C) continuous method by which LAW can be processed irrespective of whether the waste contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be comparable to LAW glass, i.e. leaches Tc-99, Re and Na at <2g/m2 during ASTM C1285 (Product Consistency) durability testing. Monolithing of the granular FBSR product was investigated to prevent dispersion during transport or burial/storage. Monolithing in an inorganic geopolymer binder, which is

  15. Fiscal Year 2010 Summary Report on the Epsilon-Metal Phase as a Waste Form for 99 Tc

    Energy Technology Data Exchange (ETDEWEB)

    Strachan, Denis M.; Crum, Jarrod V.; Buck, Edgar C.; Riley, Brian J.; Zumhoff, Mac R.

    2010-09-30

    Epsilon metal (ε-metal) is generated in nuclear fuel during irradiation. This metal consists of Pd, Ru, Rh, Mo, and some Te. These accumulate at the UO2 grain boundaries as small (ca 5 µm) particles. These metals have limited solubility in the acid used to dissolve fuel during reprocessing and in typical borosilicate glass. These must be treated separately to improve overall waste loading in glass. This low solubility and their survival in 2 Gy-old natural reactors led us to investigate them as a waste form for the immobilization of 99Tc and 107Pd, two very long-lived isotopes.

  16. Removal of cesium using coconut fiber in raw and modified forms for the treatment of radioactive liquid wastes

    Energy Technology Data Exchange (ETDEWEB)

    Jesus, Nella N.M. de; Nobre, Vanessa B.; Potiens Junior, Ademar J.; Sakata, Solange K., E-mail: sksakata@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Di Vitta, Patricia B. [Universidade de Sao Paulo (USP), SP (Brazil). Instituto de Quimica

    2013-07-01

    Sorption is one of the most studied methods to reduce the volume of radioactive waste streams. Cesium-137 is a radioisotope formed by the fission of uranium and it can cause health problems due to its easy assimilation by cells. The aim of this study is to evaluate the potential of coconut fiber in removing cesium from radioactive liquid wastes; this process can help in disposing radioactive waste. The experiments were performed in batch and the particle size of the fiber ranged between 0.30 mm and 0.50 mm. The fiber was treated with hydrogen peroxide in alkaline medium. The following parameters were analyzed: contact time, pH and concentration of cesium ions in aqueous solution. After the experiments the samples were filtered and cesium remaining in solution was quantified by inductively coupled plasma optical emission spectrometry. (author)

  17. Preliminary parametric performance assessment of potential final waste forms for alpha low-level waste at the Idaho National Engineering Laboratory. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Smith, T.H.; Sussman, M.E. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); Myers, J.; Djordjevic, S.M.; DeBiase, T.A.; Goodrich, M.T.; DeWitt, D. [IT Corp., Albuquerque, NM (United States)

    1995-08-01

    This report presents a preliminary parametric performance assessment (PA) of potential waste disposal systems for alpha-contaminated, mixed, low-level waste (ALLW) currently stored at the Transuranic Storage Area of INEL. The ALLW, which contains from 10 to 100 nCi/g of transuranic (TRU) radionuclides, is awaiting treatment and disposal. The purpose of this study was to examine the effects of several parameters on the radiological-confinement performance of potential disposal systems for the ALLW. The principal emphasis was on the performance of final waste forms (FWFs). Three categories of FWF (cement, glass, and ceramic) were addressed by evaluating the performance of two limiting FWFs for each category. Performance at five conceptual disposal sites was evaluated to illustrate the effects of site characteristics on the performance of the total disposal system. Other parameters investigated for effects on receptor dose included inventory assumptions, TRU radionuclide concentration, FWF fracture, disposal depth, water infiltration rates, subsurface-transport modeling assumptions, receptor well location, intrusion scenario assumptions, and the absence of waste immobilization. These and other factors were varied singly and in some combinations. The results indicate that compliance of the treated and disposed ALLW with the performance objectives depends on the assumptions made, as well as on the FWF and the disposal site. Some combinations result in compliance, while others do not. The implications of these results for decision making relative to treatment and disposal of the INEL ALLW are discussed. The report compares the degree of conservatism in this preliminary parametric PA against that in four other PAs and one risk assessment. All of the assessments addressed the same disposal site, but different wastes. The report also presents a qualitative evaluation of the uncertainties in the PA and makes recommendations for further study.

  18. Remaining Sites Verification Package for the 120-F-1 Glass Dump Waste Site, Waste Site Reclassification Form 2008-028

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-06-27

    The 120-F-1 waste site consisted of two dumping areas located 660 m southeast of the 105-F Reactor containing laboratory equipment and bottles, demolition debris, light bulbs and tubes, small batteries, small drums, and pesticide contaminated soil. It is probable that 108-F was the source of the debris but the material may have come from other locations within the 100-F Area. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  19. Technical Progress Report on Single Pass Flow Through Tests of Ceramic Waste Forms for Plutonium Immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, P; Roberts, S; Bourcier, W

    2000-12-01

    This report updates work on measurements of the dissolution rates of single-phase and multi-phase ceramic waste forms in flow-through reactors at Lawrence Livermore National Laboratory. Previous results were reported in Bourcier (1999). Two types of tests are in progress: (1) tests of baseline pyrochlore-based multiphase ceramics; and (2) tests of single-phase pyrochlore, zirconolite, and brannerite (the three phases that will contain most of the actinides). Tests of the multi-phase material are all being run at 25 C. The single-phase tests are being run at 25, 50, and 75 C. All tests are being performed at ambient pressure. The as-made bulk compositions of the ceramics are given in Table 1. The single pass flow-through test procedure [Knauss, 1986 No.140] allows the powdered ceramic to react with pH buffer solutions traveling upward vertically through the powder. Gentle rocking during the course of the experiment keeps the powder suspended and avoids clumping, and allows the system to behave as a continuously stirred reactor. For each test, a cell is loaded with approximately one gram of the appropriate size fraction of powdered ceramic and reacted with a buffer solution of the desired pH. The buffer solution compositions are given in Table 2. All the ceramics tested were cold pressed and sintered at 1350 C in air, except brannerite, which was sintered at 1350 C in a CO/CO{sub 2} gas mixture. They were then crushed, sieved, rinsed repeatedly in alcohol and distilled water, and the desired particle size fraction collected for the single pass flow-through tests (SPFT). The surface area of the ceramics measured by BET ranged from 0.1-0.35 m{sup 2}/g. The measured surface area values, average particle size, and sample weights for each ceramic test are given in the Appendices.

  20. Materials for Tc Capture to Increase Tc Retention in Glass Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Luksic, Steven A.; Hrma, Pavel R.; Kruger, Albert A.

    2016-04-01

    99Technetium is a long-lived fission product found in the tank waste at the Hanford site in Washington State. In its heptavalent species, it is volatile at the temperatures used in Hanford Tank Waste Treatment and Immobilization Plant vitrification melters, and thus is challenging to incorporate into waste glass. In order to decrease volatility and thereby increase retention, technetium can be converted into more thermally stable species. Several mineral phases, such as spinel, are able to incorporate tetravalent technetium in a chemically durable and thermally stable lattice, and these hosts may promote the decreased volatility that is desired. In order to be usefully implemented, there must be a synthetic rout to these phases that is compatible with both technetium chemistry and current Hanford Tank Waste Treatment and Immobilization Plant design. Synthetic routes for spinel and other potential host phases are examined.

  1. Mineral assemblage transformation of a metakaolin-based waste form after geopolymer encapsulation

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Benjamin D.; Neeway, James J.; Snyder, Michelle M. V.; Bowden, Mark E.; Amonette, James E.; Arey, Bruce W.; Pierce, Eric M.; Brown, Christopher F.; Qafoku, Nikolla P.

    2016-05-01

    Current plans for nuclear waste vitrification at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) lack the capacity to treat all of the low activity waste (LAW) that is not encapsulated in the vitrified product. Fluidized Bed Steam Reforming (FBSR) is one of the supplemental technologies under consideration to fill this gap. The FBSR process results in a granular product mainly composed of feldspathoid mineral phases that encapsulate the LAW and other contaminants of concern (COCs). In order to better understand the characteristics of the FBSR product, characterization testing has been performed on the granular product as well as the granular product encapsulated in a monolithic geopolymer binder. The non-radioactive simulated tank waste samples created for use in this study are the result of a 2008 Department of Energy sponsored Engineering Scale Technology Demonstration (ESTD) in 2008. These samples were created from waste simulant that was chemically shimmed to resemble actual tank waste, and rhenium has been used as a substitute for technetium. Another set of samples was created by the Savannah River Site Bench-Scale Reformer (BSR) using a chemical shim of Savannah River Site Tank 50 waste in order to simulate a blend of 68 Hanford tank wastes. This paper presents results from coal and moisture removal tests along with XRD, SEM, and BET analyses showing that the major mineral components are predominantly sodium aluminosilicate minerals and that the mineral product is highly porous. Results also show that the materials pass the short-term leach tests: the Toxicity Characteristic Leaching Procedure (TCLP) and Product Consistency Test (PCT).

  2. AgI-MOR Loading Effect on the Durability of the Sandia Low Temperature Sintering GCM Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Nenoff, Tina Maria [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brady, Patrick Vane. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mowry, Curtis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Garino, Terry J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-09-01

    Herein, we study the durability of the Sandia Bi-Si oxide Glass Composite Material (GCM) waste form when formulated with different weight percent levels of AgI-MOR. The post-iodine exposure AgI-MOR material was provided to SNL by ORNL. Durability results for the GCM fabricated with 22 and 25% AgI-MOR indicate releases of Ag and I at the same low rates as 15% AgI-MOR GCM, and by the same mechanism. Iodine and Ag release is controlled by the low solubility of an amorphous, hydrated silver iodide, not by the surface-controlled dissolution of I2- loaded Ag-Mordenite. Based on this data, we postulate that much higher loading levels of AgIMOR are probable in this GCM waste form, and limits will govern by retention of mechanical integrity of the GCM versus the solubility of silver iodide.

  3. Radioactive waste forms stabilized by ChemChar gasification: characterization and leaching behavior of cerium, thorium, protactinium, uranium, and neptunium.

    Science.gov (United States)

    Marrero, T W; Morris, J S; Manahan, S E

    2004-02-01

    The uses of a thermally reductive gasification process in conjunction with vitrification and cementation for the long-term disposal of low level radioactive materials have been investigated. gamma-ray spectroscopy was used for analysis of carrier-free protactinium-233 and neptunium-239 and a stoichiometric amount of cerium (observed cerium-141) subsequent to gasification and leaching, up to 48 days. High resolution ICP-MS was used to analyze the cerium, thorium, and uranium from 46 to 438 days of leaching. Leaching procedures followed the guidance of ASTM Procedure C 1220-92, Standard Test Method for Static Leaching of Monolithic Waste Forms for Disposal of Radioactive Waste. The combination of the thermally reductive pretreatment, vitrification and cementation produced a highly non-leachable form suitable for long-term disposal of cerium, thorium, protactinium, uranium, and neptunium.

  4. Investigation of microscopic radiation damage in waste forms using ODNMR and AEM techniques. 1997 annual progress report

    Energy Technology Data Exchange (ETDEWEB)

    Liu, G.

    1997-09-01

    'This project seeks to understand the microscopic effects of radiation damage in nuclear waste forms. The authors approach to this challenge encompasses studies in electron microscopy, laser spectroscopy, and computational modeling and simulation. During this first year of the project, efforts have focused on a-decay induced microscopic damage in crystalline orthophosphates (YPO{sub 4} and LuPO{sub 4}) that contain the short-lived a-emitting isotope {sup 244}Cm (t{sub 1/2} = 18.1 y). The samples that they studied were synthesized in 1980 and the initial {sup 244}Cm concentration was {approximately}1%. Studying these materials is of importance to nuclear waste management because of the opportunity to gain insight into accumulated radiation damage and the influence of aging on such damage. These factors are critical to the long-term performance of actual waste forms [1]. Lanthanide orthophosphates, including LuPO{sub 4} and YPO{sub 4}, have been suggested as waste forms for high level nuclear waste [2] and potential hosts for excess weapons plutonium [3,4]. The work is providing insight into the characteristics of these previously known radiation-resistant materials. They have observed loss of crystallinity (partial amorphization) as a direct consequence of prolonged exposure to intense alpha radiolysis in these materials. More importantly, the observation of microscopic cavities in these aged materials provides evidence of significant chemical decomposition that may be difficult to detect in the earlier stages of radiation damage. The preliminary results show that, in characterizing crystalline compounds as high level nuclear waste forms, chemical decomposition effects may be more important than lattice amorphization which has been the focus of many previous studies. More extensive studies, including in-situ analysis of the dynamics of thermal annealing of self-radiation induced amorphization and cavity formation, will be conducted on these aged {sup 244}Cm

  5. Durability of Actinide Ceramic Waste Forms Under Conditions of Granitoid Rocks

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-02-26

    Three samples of {sup 239}Pu-{sup 241}Am-doped ceramics obtained from previous research were used for alteration experiments simulating corrosion of waste forms in ion-saturated solutions. These were ceramics based on: pyrochlore, (Ca,Hf,Pu,U,Gd){sub 2}Ti{sub 2}O{sub 7}, containing 10 wt.% Pu and 0.1 wt.% Am; zircon, (Zr,Pu)SiO{sub 4}, containing 5-6 wt.% Pu and 0.05 wt.% Am; cubic zirconia, (Zr,Gd,Pu)O{sub 2}, containing 10 wt.% Pu and 0.1 wt.% Am. All these samples were milled in an agate mortar to obtain powder with particle sizes less than 30 micron. Sample of granite taken from the depth 500-503 m was studied and then used for preparing ion-saturated water solutions. A rock sample was ground, washed and classified. A fraction with particle size 0.10-0.25 mm was selected for alteration experiments. Powdered ceramic samples were separately placed into deionized water together with ground granite (approximately 1gram granite per 12-ml water) in special Teflon{trademark} vessels and set at 90 C in the oven for 3 months. After alteration experiments, the ceramic powders were studied by precise XRD analysis. Aqueous solutions and granite grains were analyzed for Am and Pu contents. The results show that alteration did not cause significant phase transformation in all ceramic samples. For all altered samples, the Am contents in aqueous solutions after experiments were similar (approximately n x 10{sup 2} Bq/ml) as well as Am amounts absorbed on granite grains (approximately n x 10{sup 5} Bq/g). Results on Pu contents were varied: for the solutions--from 60 Bq/ml for pyrochlore ceramic to 2.1 x 10{sup 3} Bq/ml for zircon ceramic; and for the absorption on granite--from 2.6 x 10{sup 4} Bq/g for zirconia ceramic to 1.4-6.8 x 10{sup 5} Bq/g for pyrochlore and zircon ceramics.

  6. Laboratory Testing of Bulk Vitrified Low-Activity Waste Forms to Support the 2005 Integrated Disposal Facility Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Eric M.; McGrail, B. Peter; Bagaasen, Larry M.; Rodriguez, Elsa A.; Wellman, Dawn M.; Geiszler, Keith N.; Baum, Steven R.; Reed, Lunde R.; Crum, Jarrod V.; Schaef, Herbert T.

    2006-06-30

    The purpose of this report is to document the results from laboratory testing of the bulk vitri-fied (BV) waste form that was conducted in support of the 2005 integrated disposal facility (IDF) performance assessment (PA). Laboratory testing provides a majority of the key input data re-quired to assess the long-term performance of the BV waste package with the STORM code. Test data from three principal methods, as described by McGrail et al. (2000a; 2003a), are dis-cussed in this testing report including the single-pass flow-through test (SPFT) and product con-sistency test (PCT). Each of these test methods focuses on different aspects of the glass corrosion process. See McGrail et al. (2000a; 2003a) for additional details regarding these test methods and their use in evaluating long-term glass performance. In addition to evaluating the long-term glass performance, this report discusses the results and methods used to provided a recommended best estimate of the soluble fraction of 99Tc that can be leached from the engineer-ing-scale BV waste package. These laboratory tests are part of a continuum of testing that is aimed at improving the performance of the BV waste package.

  7. Laboratory Testing of Bulk Vitrified Low-Activity Waste Forms to Support the 2005 Integrated Disposal Facility Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Eric M.; McGrail, B. Peter; Bagaasen, Larry M.; Rodriguez, Elsa A.; Wellman, Dawn M.; Geiszler, Keith N.; Baum, Steven R.; Reed, Lunde R.; Crum, Jarrod V.; Schaef, Herbert T.

    2005-03-31

    The purpose of this report is to document the results from laboratory testing of the bulk vitri-fied (BV) waste form that was conducted in support of the 2005 integrated disposal facility (IDF) performance assessment (PA). Laboratory testing provides a majority of the key input data re-quired to assess the long-term performance of the BV waste package with the STORM code. Test data from three principal methods, as described by McGrail et al. (2000a; 2003a), are dis-cussed in this testing report including the single-pass flow-through test (SPFT) and product con-sistency test (PCT). Each of these test methods focuses on different aspects of the glass corrosion process. See McGrail et al. (2000a; 2003a) for additional details regarding these test methods and their use in evaluating long-term glass performance. In addition to evaluating the long-term glass performance, this report discusses the results and methods used to provided a recommended best estimate of the soluble fraction of 99Tc that can be leached from the engineer-ing-scale BV waste package. These laboratory tests are part of a continuum of testing that is aimed at improving the performance of the BV waste package.

  8. CORRELATION OF POLYCHLORINATED NAPHTHALENES WITH POLYCHLORINATED DIBENZOFURANS FORMED FROM WASTE INCINERATION

    Science.gov (United States)

    Isomer composition of polychlorinated naphthalenes (PCNs) was measured for municipal waste incinerator fly ash samples,and for emission samples produced from soot and copper deposit experiments conducted at EPA. Two types of PCN isomer patterns were identified. One pattern cxonta...

  9. Irradiation effect on leaching behavior and form of heavy metals in fly ash of municipal solid waste incinerator.

    Science.gov (United States)

    Nam, Sangchul; Namkoong, Wan

    2012-01-15

    Fly ash from a municipal solid waste incinerator (MSWI) is commonly classified as hazardous waste. High-energy electron beam irradiation systems have gained popularity recently as a clean and promising technology to remove environmental pollutants. Irradiation effects on leaching behavior and form of heavy metals in MSWI fly ash have not been investigated in any significant detail. An electron beam accelerator was used in this research. Electron beam irradiation on fly ash significantly increased the leaching potential of heavy metals from fly ash. The amount of absorbed dose and the metal species affected leaching behavior. When electron beam irradiation intensity increased gradually up to 210 kGy, concentration of Pb and Zn in the leachate increased linearly as absorbed dose increased, while that of Cu underwent no significant change. Concentration of Pb and Zn in the leachate increased up to 15.5% (10.7 mg/kg), and 35.6% (9.6 mg/kg) respectively. However, only 4.8% (0.3mg/kg) increase was observed in the case of Cu. The results imply that irradiation has significant effect on the leaching behavior of heavy metals in fly ash, and the effect is quite different among the metal species tested in this study. A commonly used sequential extraction analysis which can classify a metal species into five forms was conducted to examine any change in metal form in the irradiated fly ash. Notable change in metal form in fly ash was observed when fly ash was irradiated. Change in Pb form was much greater than that of Cu form. Change in metal form was related to leaching potential of the metals. Concentration of heavy metal in leachate was positively related to the exchangeable form which is the most mobile. It may be feasible to treat fly ash by electron beam irradiation for selective recovery of valuable metals or for pretreatment prior to conventional processes.

  10. Five-Year Implementation Plan For Advanced Separations and Waste Forms Capabilities at the Idaho National Laboratory (FY 2011 to FY 2015)

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2011-03-01

    DOE-NE separations research is focused today on developing a science-based understanding that builds on historical research and focuses on combining a fundamental understanding of separations and waste forms processes with small-scale experimentation coupled with modeling and simulation. The result of this approach is the development of a predictive capability that supports evaluation of separations and waste forms technologies. The specific suite of technologies explored will depend on and must be integrated with the fuel development effort, as well as an understanding of potential waste form requirements. This five-year implementation plan lays out the specific near-term tactical investments in people, equipment and facilities, and customer capture efforts that will be required over the next five years to quickly and safely bring on line the capabilities needed to support the science-based goals and objectives of INL’s Advanced Separations and Waste Forms RD&D Capabilities Strategic Plan.

  11. New data on mineral forms of rare metals in phosphogypsum wastes

    Science.gov (United States)

    Samonov, A. E.

    2011-09-01

    Phosphogypsum is an industrial waste of the processing of Khibiny apatite concentrate into chemical fertilizers by sulfurous technology. This is a valuable and promising technogenous rare-metal feedstock. The samples of fresh and old phosphogypsum were studied using precision physical techniques of analytical electron microscopy and X-ray spectral microanalysis. These studies allowed the discovery of new and unusual mineral compositions including strontium and rare earth metals in mineral fractions of phosphogypsum. The appearance of a new generation of technogenous rare-metal raw material permits us to characterize the prospects of its industrial use and to develop nonwaste technologies of its complex treatment.

  12. Laboratory Testing of Bulk Vitrified Low-Activity Waste Forms to Support the 2005 Integrated Disposal Facility Performance Assessment Erratum

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-06

    This report refers to or contains Kg values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected Kg values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

  13. Laboratory Testing of Bulk Vitrified Low-Activity Waste Forms to Support the 2005 Integrated Disposal Facility Performance Assessment. Erratum

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-06

    This report refers to or contains Kg values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected Kg values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

  14. Candidate Low-Temperature Glass Waste Forms for Technetium-99 Recovered from Hanford Effluent Management Facility Evaporator Concentrate

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Mei [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tang, Ming [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rim, Jung Ho [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chamberlin, Rebecca M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-07-24

    Alternative treatment and disposition options may exist for technetium-99 (99Tc) in secondary liquid waste from the Hanford Direct-Feed Low-Activity Waste (DFLAW) process. One approach includes development of an alternate glass waste form that is suitable for on-site disposition of technetium, including salts and other species recovered by ion exchange or precipitation from the EMF evaporator concentrate. By recovering the Tc content from the stream, and not recycling the treated concentrate, the DFLAW process can potentially be operated in a more efficient manner that lowers the cost to the Department of Energy. This report provides a survey of candidate glass formulations and glass-making processes that can potentially incorporate technetium at temperatures <700 °C to avoid volatilization. Three candidate technetium feed streams are considered: (1) dilute sodium pertechnetate loaded on a non-elutable ion exchange resin; (2) dilute sodium-bearing aqueous eluent from ion exchange recovery of pertechnetate, or (3) technetium(IV) oxide precipitate containing Sn and Cr solids in an aqueous slurry. From the technical literature, promising candidate glasses are identified based on their processing temperatures and chemical durability data. The suitability and technical risk of three low-temperature glass processing routes (vitrification, encapsulation by sintering into a glass composite material, and sol-gel chemical condensation) for the three waste streams was assessed, based on available low-temperature glass data. For a subset of candidate glasses, their long-term thermodynamic behavior with exposure to water and oxygen was modeled using Geochemist’s Workbench, with and without addition of reducing stannous ion. For further evaluation and development, encapsulation of precipitated TcO2/Sn/Cr in a glass composite material based on lead-free sealing glasses is recommended as a high priority. Vitrification of pertechnetate in aqueous anion exchange eluent solution

  15. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 3A, ORIGEN2 decay tables for immobilized high-level waste, Appendix 3B, Interim high-level waste forms, Appendix 3C, User's guide to the high-level waste PC data base

    Energy Technology Data Exchange (ETDEWEB)

    1987-12-01

    The purpose of this report, and the information contained in the associated computerized data bases, is to establish the DOE/OCRWM reference characteristics of the radioactive waste materials that may be accepted by DOE for emplacement in he mined geologic disposal system. This report provides relevant technical data for use by DOE and its supporting contractors and is not intended to be a policy document. This document is backed up by five PC-compatible data bases, written in a user-oriented, menu-driven format, which were developed for this purpose. The data bases are the LWR Assemblies Data Base; the LWR Radiological Data Base; the LWR Quantities Data Base; the LWR NFA Hardware Data Base; and the High-Level Waste Data Base. The above data bases may be ordered using the included form. Volume 6 contains decay tables for immobilized high-level waste, information on interim high-level waste forms, and a user's guide to the high-level waste PC data base.

  16. Physical, Chemical and Structural Evolution of Zeolite-Containing Waste Forms Produced from Metakaolinite and Calcined Sodium Bearing Waste (HLW and/or LLW)

    Energy Technology Data Exchange (ETDEWEB)

    Grutzeck, Michael W.

    2005-06-27

    Zeolites are extremely versatile. They can adsorb liquids and gases and serve as cation exchange media. They occur in nature as well cemented deposits. The ancient Romans used blocks of zeolitized tuff as a building material. Using zeolites for the management of radioactive waste is not a new idea, but a process by which the zeolites can be made to act as a cementing agent is. Zeolitic materials are relatively easy to synthesize from a wide range of both natural and man-made substances. The process under study is derived from a well known method in which metakaolin (an impure thermally dehydroxylated kaolinite heated to {approx}700 C containing traces of quartz and mica) is mixed with sodium hydroxide (NaOH) and reacted in slurry form (for a day or two) at mildly elevated temperatures. The zeolites form as finely divided powders containing micrometer ({micro}m) sized crystals. However, if the process is changed slightly and only just enough concentrated sodium hydroxide solution is added to the metakaolinite to make a thick crumbly paste and then the paste is compacted and cured under mild hydrothermal conditions (60-200 C), the mixture will form a hard ceramic-like material containing distinct crystalline tectosilicate minerals (zeolites and feldspathoids) imbedded in an X-ray amorphous hydrated sodium aluminosilicate matrix. Due to its lack of porosity and vitreous appearance we have chosen to call this composite a ''hydroceramic''.

  17. Prediction models of long-term leaching behavior and leaching mechanism of glass components and surrogated nuclides in radioactive vitrified waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Y. C.; Lee, K. S. [Department of Industrial Environment and Health, Yonsei University, Wonju (Korea, Republic of); Kim, I. T.; Kim, H. T.; Kim, J. H. [Korea Atomic Energy Research Institute (KAERI), Taejon (Korea, Republic of)

    1999-07-01

    Melting solidification is considered to be a perspective technology for stabilizing incineration ash remaining after incineration of combustible radioactive waste since it has the advantage of improving the physicochemical properties of waste forms. Final waste forms should be characterized to determine the degree to which they fulfills the acceptance criteria of the disposal facility. Chemical durability (leaching resistance) is known to be the most important factor in the assessment of waste forms. In this study, vitrified waste forms are manufactured and characterized. Feed materials consist of simulated radioactive incineration ash and base-glass with different mixing ratios. To assess the chemical durability of vitrified waste forms, the International Standard Organization (ISO) leach test has been conducted at 70 degree C with deionized distilled water as a leachant for 820 days, and the concentrations of glass components and surrogates in the leachates are then analyzed. Two models for predicting long-term leaching behavior of glass components and radionuclides in a glass form are applied to the leached data after 820 days. The model including a fitted parameter from the longer experimental data shows more accuracy, however, the model with shorter leaching test results offers the advantage of being able to predict the long-term behavior from the short-term experimental data. The leaching mechanisms of surrogates and glass components were also investigated by using two semi-empirical kinetic models and were found to be dissolution with diffusion. (author)

  18. Standard practice for prediction of the long-term behavior of materials, including waste forms, used in engineered barrier systems (EBS) for geological disposal of high-level radioactive waste

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice describes test methods and data analyses used to develop models for the prediction of the long-term behavior of materials, such as engineered barrier system (EBS) materials and waste forms, used in the geologic disposal of spent nuclear fuel (SNF) and other high-level nuclear waste in a geologic repository. The alteration behavior of waste form and EBS materials is important because it affects the retention of radionuclides by the disposal system. The waste form and EBS materials provide a barrier to release either directly (as in the case of waste forms in which the radionuclides are initially immobilized), or indirectly (as in the case of containment materials that restrict the ingress of groundwater or the egress of radionuclides that are released as the waste forms and EBS materials degrade). 1.1.1 Steps involved in making such predictions include problem definition, testing, modeling, and model confirmation. 1.1.2 The predictions are based on models derived from theoretical considerat...

  19. Photoproducts of carminic acid formed by a composite from Manihot dulcis waste.

    Science.gov (United States)

    Antonio-Cisneros, Cynthia M; Dávila-Jiménez, Martín M; Elizalde-González, María P; García-Díaz, Esmeralda

    2015-04-15

    Carbon-TiO2 composites were obtained from carbonised Manihot dulcis waste and TiO2 using glycerol as an additive and thermally treating the composites at 800 °C. Furthermore, carbon was obtained from manihot to study the adsorption, desorption and photocatalysis of carminic acid on these materials. Carminic acid, a natural dye extracted from cochineal insects, is a pollutant produced by the food industry and handicrafts. Its photocatalysis was observed under different atmospheres, and kinetic curves were measured by both UV-Vis and HPLC for comparison, yielding interesting differences. The composite was capable of decomposing approximately 50% of the carminic acid under various conditions. The reaction was monitored by UV-Vis spectroscopy and LC-ESI-(Qq)-TOF-MS-DAD, enabling the identification of some intermediate species. The deleterious compound anthracene-9,10-dione was detected both in N2 and air atmospheres.

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

    Science.gov (United States)

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

    2009-11-15

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

  1. Prototype Development of Remote Operated Hot Uniaxial Press (ROHUP) to Fabricate Advanced Tc-99 Bearing Ceramic Waste Forms - 13381

    Energy Technology Data Exchange (ETDEWEB)

    Alaniz, Ariana J.; Delgado, Luc R.; Werbick, Brett M. [University of Nevada - Las Vegas, Howard R. Hughes College of Engineering, 4505 S. Maryland Parkway, Box 454009, Las Vegas, NV 89154-4009 (United States); Hartmann, Thomas [University of Nevada - Las Vegas, Harry Reid Canter, 4505 S. Maryland Parkway, Box 454009, Las Vegas, NV 89154-4009 (United States)

    2013-07-01

    The objective of this senior student project is to design and build a prototype construction of a machine that simultaneously provides the proper pressure and temperature parameters to sinter ceramic powders in-situ to create pellets of rather high densities of above 90% (theoretical). This ROHUP (Remote Operated Hot Uniaxial Press) device is designed specifically to fabricate advanced ceramic Tc-99 bearing waste forms and therefore radiological barriers have been included in the system. The HUP features electronic control and feedback systems to set and monitor pressure, load, and temperature parameters. This device operates wirelessly via portable computer using Bluetooth{sup R} technology. The HUP device is designed to fit in a standard atmosphere controlled glove box to further allow sintering under inert conditions (e.g. under Ar, He, N{sub 2}). This will further allow utilizing this HUP for other potential applications, including radioactive samples, novel ceramic waste forms, advanced oxide fuels, air-sensitive samples, metallic systems, advanced powder metallurgy, diffusion experiments and more. (authors)

  2. Enhanced 99 Tc retention in glass waste form using Tc(IV)-incorporated Fe minerals

    Energy Technology Data Exchange (ETDEWEB)

    Um, Wooyong; Luksic, Steven A.; Wang, Guohui; Saslow, Sarah; Kim, Dong-Sang; Schweiger, Michael J.; Soderquist, Chuck Z.; Bowden, Mark E.; Lukens, Wayne W.; Kruger, Albert A.

    2017-11-01

    Technetium (99Tc) immobilization by doping into iron oxide mineral phases may alleviate the problems with Tc volatility during vitrification of nuclear waste. Reduced Tc, Tc(IV), substitutes for Fe(III) in the crystal structure by a process of Tc reduction from Tc(VII) to Tc(IV) followed by co-precipitation of Fe oxide minerals. Two Tc-incorporated Fe minerals (Tc-goethite and Tc-magnetite/maghemite) were prepared and tested for Tc retention in glass melt samples at temperatures between 600 – 1,000 oC. After being cooled, the solid glass specimens prepared at different temperatures were analyzed for Tc oxidation state using Tc K-edge XANES. In most samples, Tc was partially oxidized from Tc(IV) to Tc(VII) as the melt temperature increased. However, Tc retention in glass melt samples prepared using Tc-incorporated Fe minerals were moderately higher than in glass prepared using KTcO4 because of limited and delayed Tc volatilization.

  3. 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.

  4. Process simulation and statistical approaches for validating waste form qualification models

    Energy Technology Data Exchange (ETDEWEB)

    Kuhn, W.L.; Toland, M.R.; Pulsipher, B.A.

    1989-05-01

    This report describes recent progress toward one of the principal objectives of the Nuclear Waste Treatment Program (NWTP) at the Pacific Northwest Laboratory (PNL): to establish relationships between vitrification process control and glass product quality. during testing of a vitrification system, it is important to show that departures affecting the product quality can be sufficiently detected through process measurements to prevent an unacceptable canister from being produced. Meeting this goal is a practical definition of a successful sampling, data analysis, and process control strategy. A simulation model has been developed and preliminarily tested by applying it to approximate operation of the West Valley Demonstration Project (WVDP) vitrification system at West Valley, New York. Multivariate statistical techniques have been identified and described that can be applied to analyze large sets of process measurements. Information on components, tanks, and time is then combined to create a single statistic through which all of the information can be used at once to determine whether the process has shifted away from a normal condition.

  5. Materials Characterization Center workshop on leaching mechanisms of nuclear waste forms, May 19-21, 1982, Gaithersburg, Maryland. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Mendel, J.E. (comp.)

    1982-08-01

    This is a report of the second workshop on the leaching mechanism of nuclear waste forms, which was held at Geithersburg, Maryland, May 19-21, 1982. The first session of the workshop was devoted to progress reports by participants in the leaching mechanisms program. These progress reports, as prepared by the participants, are given in Section 3.0. The goal of the remainder of the workshop was to exchange information on the development of repository-relevant leach testing techniques, often called interactions testing. To this end, a wide spectrum of investigators, many of whose work is sponsored by DOE's Nuclear Waste Terminal Storage (NWTS) project, made presentations at the workshop. These presentations were a significant and beneficial part of the workshop and are summarized in Sections 4.0, 5.0 and 6.0 according to the workshop agenda topics. In many cases, the presenters provided a written version of their presentation which has been included verbatim; in the other cases, the workshop chairman has supplied a brief synopsis. Twenty-one papers have been abstracted and indexed for inclusion in the data base.

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

    Science.gov (United States)

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

    2015-11-17

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

  7. Production of Welding Fluxes Using Waste Slag Formed in Silicomanganese Smelting

    Science.gov (United States)

    Kozyrev, N. A.; Kryukov, R. E.; Kozyreva, O. E.; Lipatova, U. I.; Filonov, A. V.

    2016-04-01

    The possibility in principle of using slag, which is formed in the silicon-manganese smelting process, in producing welding fluxes is shown. The composition of and technology used for a new fused flux has been designed. A comparative evaluation of the new flux and the widely used AN-348 type flux was done. It has been proved that the new flux has high strength properties.

  8. Standard test method for accelerated leach test for diffusive releases from solidified waste and a computer program to model diffusive, fractional leaching from cylindrical waste forms

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This test method provides procedures for measuring the leach rates of elements from a solidified matrix material, determining if the releases are controlled by mass diffusion, computing values of diffusion constants based on models, and verifying projected long-term diffusive releases. This test method is applicable to any material that does not degrade or deform during the test. 1.1.1 If mass diffusion is the dominant step in the leaching mechanism, then the results of this test can be used to calculate diffusion coefficients using mathematical diffusion models. A computer program developed for that purpose is available as a companion to this test method (Note 1). 1.1.2 It should be verified that leaching is controlled by diffusion by a means other than analysis of the leach test solution data. Analysis of concentration profiles of species of interest near the surface of the solid waste form after the test is recommended for this purpose. 1.1.3 Potential effects of partitioning on the test results can...

  9. Influence of fracture networks on radionuclide transport from solidified waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Seetharam, S.C., E-mail: suresh.seetharam@sckcen.be [Performance Assessments Unit, Institute for Environment, Health and Safety, Belgian Nuclear Research Centre (SCK CEN), Boeretang 200, B-2400 Mol (Belgium); Perko, J.; Jacques, D. [Performance Assessments Unit, Institute for Environment, Health and Safety, Belgian Nuclear Research Centre (SCK CEN), Boeretang 200, B-2400 Mol (Belgium); Mallants, D. [CSIRO Land and Water, Waite Road – Gate 4, Glen Osmond, SA 5064 (Australia)

    2014-04-01

    Highlights: • Magnitude of peak radionuclide fluxes is less sensitive to the fracture network geometry. • Time of peak radionuclide fluxes is sensitive to the fracture networks. • Uniform flow model mimics a limiting case of a porous medium with large number of fine fractures. • Effect of fracture width on radionuclide flux depends on the ratio of fracture to matrix conductivity. • Effect of increased dispersivity in fractured media does not always result in a lower peak flux for specific fracture networks due to higher concentrations adjacent to the fracture plane. - Abstract: Analysis of the effect of fractures in porous media on fluid flow and mass transport is of great interest in many fields including geotechnical, petroleum, hydrogeology and waste management. This paper presents sensitivity analyses examining the effect of various hypothetical fracture networks on the performance of a planned near surface disposal facility in terms of radionuclide transport behaviour. As it is impossible to predict the initiation and evolution of fracture networks and their characteristics in concrete structures over time scales of interest, several fracture networks have been postulated to test the sensitivity of radionuclide release from a disposal facility. Fluid flow through concrete matrix and fracture networks are modelled via Darcy's law. A single species radionuclide transport equation is employed for both matrix and fracture networks, which include the processes advection, diffusion, dispersion, sorption/desorption and radioactive decay. The sensitivity study evaluates variations in fracture network configuration and fracture width together with different sorption/desorption characteristics of radionuclides in a cement matrix, radioactive decay constants and matrix dispersivity. The effect of the fractures is illustrated via radionuclide breakthrough curves, magnitude and time of peak mass flux, cumulative mass flux and concentration profiles. For the

  10. Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain reporsitory.

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, W. E.

    2006-01-31

    The results of laboratory testing and modeling activities conducted to support the development of waste forms to immobilize wastes generated during the electrometallurgical treatment of spent sodium-bonded nuclear fuel and their qualification for disposal in the federal high-level radioactive waste repository are summarized in this report. Tests and analyses were conducted to address issues related to the chemical, physical, and radiological properties of the waste forms relevant to qualification. These include the effects of composition and thermal treatments on the phase stability, radiation effects, and methods for monitoring product consistency. Other tests were conducted to characterize the degradation and radionuclide release behaviors of the ceramic waste form (CWF) used to immobilize waste salt and the metallic waste form (MWF) used to immobilize metallic wastes and to develop models for calculating the release of radionuclides over long times under repository-relevant conditions. Most radionuclides are contained in the binder glass phase of the CWF and in the intermetallic phase of the MWF. The release of radionuclides from the CWF is controlled by the dissolution rate of the binder glass, which can be tracked using the same degradation model that is used for high-level radioactive waste (HLW) glass. Model parameters measured for the aqueous dissolution of the binder glass are used to model the release of radionuclides from a CWF under all water-contact conditions. The release of radionuclides from the MWF is element-specific, but the release of U occurs the fastest under most test conditions. The fastest released constituent was used to represent all radionuclides in model development. An empirical aqueous degradation model was developed to describe the dependence of the radionuclide release rate from a MWF on time, pH, temperature, and the Cl{sup -} concentration. The models for radionuclide release from the CWF and MWF are both bounded by the HLW glass

  11. Utilization of orange peel, a food industrial waste, in the production of exo-polygalacturonase by pellet forming Aspergillus sojae.

    Science.gov (United States)

    Buyukkileci, Ali Oguz; Lahore, Marcello Fernandez; Tari, Canan

    2015-04-01

    The production of exo-polygalacturonase (exo-PG) from orange peel (OP), a food industrial waste, using Aspergillus sojae was studied in submerged culture. A simple, low-cost, industrially significant medium formulation, composed of only OP and (NH4)2SO4 (AS) was developed. At an inoculum size of 2.8 × 10(3) spores/mL, growth was in the form of pellets, which provided better mixing of the culture broth and higher exo-PG activity. These pellets were successfully used as an inoculum for bioreactors and 173.0 U/mL exo-PG was produced. Fed-batch cultivation further enhanced the exo-PG activity to 244.0 U/mL in 127.5 h. The final morphology in the form of pellets is significant to industrial fermentation easing the subsequent downstream processing. Furthermore, the low pH trend obtained during this fermentation serves an advantage to fungal fermentations prone to contamination problems. As a result, an economical exo-PG production process was defined utilizing a food industrial by-product and producing high amount of enzyme.

  12. Yucca Mountain project : FY 2006 annual report for waste form testingactivities.

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, W. L.; Fortner, J. A.; Guelis, A. V.; Cunnane, J. C.

    2006-11-01

    This report describes the experimental work performed at Argonne National Laboratory (Argonne) during fiscal year (FY 2006) under the Bechtel SAIC Company, LLC (BSC) Memorandum Purchase Order (MPO) contract number B004210CM3X. Because this experimental work is focused on the dissolution and precipitation behavior of neptunium, the report also includes, or incorporates by reference, earlier results that are relevant to presenting a more complete understanding of the likely behavior of neptunium under experimental conditions relevant to the Yucca Mountain repository. Important results relevant to the technical bases, validations, and conservatisms in current source term models are summarized. The CSNF samples were observed to corrode following the general contour of the surface rather than via (for instance) grain boundary attack. This supports the current approach of estimating the effective surface area of corroding CSNF based on the geometric surface area of fuel pellet fragments. It was observed that the neptunium and plutonium concentrations in corroded CSNF samples were somewhat higher at and near the corrosion front (i.e., at the interface between the alteration product ''rind'' layer and the underlying fuel) than in the bulk fuel. The neptunium and plutonium at the corrosion front and in the uranyl alteration layer were found to be in the quadravalent (4+) oxidation state. The uranyl phases that constitute most of the alteration rind were depleted in neptunium relative to the bulk fuel: neptunium concentrations in the uranyl alteration rind were less than 20% of that in the parent fuel. Homogeneous precipitation tests have shown that solids precipitate from a 1 x 10{sup -4} M Np(V) solution over the temperature range of 200-280 C, but no evidence was found that any solids precipitated from the same solution at 150 C through 289 days. The solids formed in the homogeneous precipitation tests were predominantly a Np(IV)-bearing phase

  13. Hanford facility dangerous waste Part A, Form 3 and Part B permit application documentation, Central Waste Complex (WA7890008967)(TSD: TS-2-4)

    Energy Technology Data Exchange (ETDEWEB)

    Saueressig, D.G.

    1998-05-20

    The Hanford Facility Dangerous Waste Permit Application is considered to be a single application organized into a General Information Portion (document number DOE/RL-91-28) and a Unit-Specific Portion. The scope of the Unit-Specific Portion is limited to Part B permit application documentation submitted for individual, operating, treatment, storage, and/or disposal units, such as the Central Waste Complex (this document, DOE/RL-91-17). Both the General Information and Unit-Specific portions of the Hanford Facility Dangerous Waste Permit Application address the content of the Part B permit application guidance prepared by the Washington State Department of Ecology (Ecology 1996) and the U.S. Environmental Protection Agency (40 Code of Federal Regulations 270), with additional information needed by the Hazardous and Solid Waste Amendments and revisions of Washington Administrative Code 173-303. For ease of reference, the Washington State Department of Ecology alpha-numeric section identifiers from the permit application guidance documentation (Ecology 1996) follow, in brackets, the chapter headings and subheadings. A checklist indicating where information is contained in the Central Waste Complex permit application documentation, in relation to the Washington State Department of Ecology guidance, is located in the Contents section. Documentation contained in the General Information Portion is broader in nature and could be used by multiple treatment, storage, and/or disposal units (e.g., the glossary provided in the General Information Portion). Wherever appropriate, the Central Waste Complex permit application documentation makes cross-reference to the General Information Portion, rather than duplicating text. Information provided in this Central Waste Complex permit application documentation is current as of May 1998.

  14. Remaining Sites Verification Package for the 116-C-3, 105-C Chemical Waste Tanks, Waste Site Reclassification Form 2008-002

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-01-31

    The 116-C-3 waste site consisted of two underground storage tanks designed to receive mixed waste from the 105-C Reactor Metals Examination Facility chemical dejacketing process. Confirmatory evaluation and subsequent characterization of the site determined that the southern tank contained approximately 34,000 L (9,000 gal) of dejacketing wastes, and that the northern tank was unused. In accordance with this evaluation, the verification sampling and modeling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrate that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also show that residual contaminant concentrations are protective of groundwater and the Columbia River.

  15. Interim Report on Development of a Model to Predict Dissolution Behavior of the Titanate Waste Form in a Repository

    Energy Technology Data Exchange (ETDEWEB)

    Bourcier, W.L.

    1999-08-16

    Dissolution testing performed to date on a titanate waste form under development for plutonium immobilization reveals the following: (1) The wasteform is very durable. Many of the test results have shown the dissolution rate to be below detection or less than background levels of the constituent elements; (2) elemental release is non-stoichiometric with Pu, U, Ca, and Gd released faster than Ti and Hf at most pH conditions; (3) dissolution rates measured in flow-through tests sometimes show a continuous decrease with time in tests of up to two years duration; (4) attempts to model the dissolution as a transport-controlled process with diffusion through a leached layer as the rate limiting mechanism show reasonable agreement at low pH conditions but poor agreement at neutral to alkaline pHs. Based on present uncertainties in our understanding of rate control, we have provided conservative estimates of radionuclide release rates based on the fastest observed release rates measured in short-term tests. These dissolution rates under repository-relevant conditions are in the range of 10{sup -3} to 10{sup -6}g/m{sup 2}/day.

  16. Enhancing the growth and yield of Ramie (Boehmeria nivea L.) by ramie biomass waste in liquid form and gibberellic acid

    Science.gov (United States)

    Suherman, C.; Nuraini, A.; Wulandari, A. P.; Kadapi, M.

    2017-05-01

    Ramie (Boehmeria nivea L.) is one of the most important sources of natural fibre, a sustainable biomass. The growth and yield of ramie are affected by mineral nutrients. In the present study, we usedfertilizers from waste of ramie biomass in liquid form (liquid organic fertilizer, LOF) and the other treatment is by gibberellic acid (GA3). This study was to obtain the effect of treatments on enhance the growth and yield of ramie. Hence, we measure the character that related to the important parameter for biomass product of ramie. Such plant height, stem diameter, dry plant weight, and ramie fresh stem weight of ramie clone Pujon 13. This research was conducted from January 2016 to March 2016 at Research Field Ciparanje, Faculty of Agriculture, Padjadjaran University, Jatinangor, Sumedang, West Java with an altitude of about ± 750 m above sea level. The type of Soil in this area is Inceptisolsoil order and thetype of rainfall according to Schmidt and Fergusson Classification is C type. The experiment used Randomized Block Design (RBD) which consisted of eight treatments (GA and LOF) and four replications. The concentration of GA from 0, 50, 100 and 150 ppm and for concentration of LOF is 40 mlL-1. We suggested the treatment of GA 150 ppm with 40 mlL-1 LOF was the best treatment on enhancing plant height and stem fresh weight of ramie clone Pujon 13.

  17. Remaining Sites Verification Package for the 100-B-1 Surface Chemical and Solid Waste Dumping Area, Waste Site Reclassification Form 2006-003

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Carlson

    2006-04-24

    The 100-B-1 waste site was a dumping site that was divided into two areas. One area was used as a laydown area for construction materials, and the other area was used as a chemical dumping area. The 100-B-1 Surface Chemical and Solid Waste Dumping Area site meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  18. Experimental determination of the speciation, partitioning, and release of perrhenate as a chemical surrogate for pertechnetate from a sodalite-bearing multiphase ceramic waste form

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Eric M.; Lukens, Wayne W.; Fitts, Jeff. P.; Jantzen, Carol. M.; Tang, G.

    2013-12-01

    A key component to closing the nuclear fuel cycle is the storage and disposition of nuclear waste in geologic systems. Multiphase ceramic waste forms have been studied extensively as a potential host matrix for nuclear waste. Understanding the speciation, partitioning, and release behavior of radionuclides immobilized in multiphase ceramic waste forms is a critical aspect of developing the scientific and technical basis for nuclear waste management. In this study, we evaluated a sodalite-bearing multiphase ceramic waste form (i.e., fluidized-bed steam reform sodium aluminosilicate [FBSR NAS] product) as a potential host matrix for long-lived radionuclides, such as technetium (99Tc). The FBSR NAS material consists primarily of nepheline (ideally NaAlSiO4), anion-bearing sodalites (ideally M8[Al6Si6O24]X2, where M refers to alkali and alkaline earth cations and X refers to monovalent anions), and nosean (ideally Na8[AlSiO4]6SO4). Bulk X-ray absorption fine structure analysis of the multiphase ceramic waste form, suggest rhenium (Re) is in the Re(VII) oxidation state and has partitioned to a Re-bearing sodalite phase (most likely a perrhenate sodalite Na8[Al6Si6O24](ReO4)2). Rhenium was added as a chemical surrogate for 99Tc during the FBSR NAS synthesis process. The weathering behavior of the FBSR NAS material was evaluated under hydraulically unsaturated conditions with deionized water at 90 ?C. The steady-state Al, Na, and Si concentrations suggests the weathering mechanisms are consistent with what has been observed for other aluminosilicate minerals and include a combination of ion exchange, network hydrolysis, and the formation of an enriched-silica surface layer or phase. The steady-state S and Re concentrations are within an order of magnitude of the nosean and perrhenate sodalite solubility, respectively. The order of magnitude difference between the observed and predicted concentration for Re and S may be associated with the fact that the anion

  19. Experimental Determination of the Speciation, Partitioning, and Release of Perrhenate as a Chemical Surrogate for Pertechnetate from a Sodalite-Bearing Multiphase Ceramic Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Eric M [ORNL; Lukens, Wayne W [Lawrence Berkeley National Laboratory (LBNL); Fitts, Jeffrey P [Princeton University; Tang, Guoping [ORNL; Jantzen, C M [Savannah River National Laboratory (SRNL)

    2013-01-01

    A key component to closing the nuclear fuel cycle is the storage and disposition of nuclear waste in geologic systems. Multiphase ceramic waste forms have been studied extensively as a potential host matrix for nuclear waste. Understanding the speciation, partitioning, and release behavior of radionuclides immobilized in multiphase ceramic waste forms is a critical aspect of developing the scientific and technical basis for nuclear waste management. In this study, we evaluated a sodalite-bearing multiphase ceramic waste form (i.e., fluidized-bed steam reform sodium aluminosilicate [FBSR NAS] product) as a potential host matrix for long-lived radionuclides, such as technetium (99Tc). The FBSR NAS material consists primarily of nepheline (ideally NaAlSiO4), anion-bearing sodalites (ideally M8[Al6Si6O24]X2, where M refers to alkali and alkaline earth cations and X refers to monovalent anions), and nosean (ideally Na8[AlSiO4]6SO4). Bulk x-ray absorption fine structure analysis of the multiphase ceramic waste form, suggest rhenium (Re) is in the Re(VII) oxidation state and has partitioned to a Re-bearing sodalite phase (most likely a perrhenate sodalite Na8[Al6Si6O24](ReO4)2). Rhenium was added as a chemical surrogate for 99Tc during the FBSR NAS synthesis process. The weathering behavior of the FBSR NAS material was evaluated under hydraulically unsaturated conditions with deionized water at 90 C. The steady-state Al, Na, and Si concentrations suggests the weathering mechanisms are consistent with what has been observed for other aluminosilicate minerals and include a combination of ion exchange, network hydrolysis, and the formation of an enriched-silica surface layer or phase. The steady-state S and Re concentrations are within an order of magnitude of the nosean and perrhenate sodalite solubility, respectively. The order of magnitude difference between the observed and predicted concentration for Re and S may be associated with the fact that the anion

  20. Remaining Sites Verification Package for the 100-B-23, 100-B/C Area Surface Debris, Waste Site, Waste Site Reclassification Form 2008-027

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-06-16

    The 100-B-23, 100-B/C Surface Debris, waste consisted of multiple locations of surface debris and chemical stains that were identified during an Orphan Site Evaluation of the 100-B/C Area. Evaluation of the collected information for the surface debris features yielded four generic waste groupings: asbestos-containing material, lead debris, oil and oil filters, and treated wood. Focused verification sampling was performed concurrently with remediation. Site remediation was accomplished by selective removal of the suspect hazardous items and potentially impacted soils. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  1. Informative document waste plastics

    NARCIS (Netherlands)

    Nagelhout D; Sein AA; Duvoort GL

    1989-01-01

    This "Informative document waste plastics" forms part of a series of "informative documents waste materials". These documents are conducted by RIVM on the indstruction of the Directorate General for the Environment, Waste Materials Directorate, in behalf of the program of

  2. Microwave irradiated synthesis and characterization of 1, 4-phenylene bis-oxazoline form bis-(2-hydroxyethyl) terephthalamide obtained by depolymerization of poly (ethylene terephthalate) (PET) bottle wastes

    OpenAIRE

    Yogesh S. Parab; Rikhil V. Shah; Sanjeev R. Shukla

    2012-01-01

    The aminolytic depolymerization of PET bottle waste with ethanolamine by conventional heating and microwave irradiation heating method was attempted with heterogeneous, recyclable acid catalysts such as beta zeolite (SiO2/ AlO2= 15 Na- form) and montmorillonite KSF. The pure product bis-(2-hydroxyethyl) terephthalamide (BHETA) of aminolysis was obtained in good yield (85- 88%). The BHETA, thus obtained, was subjected to cyclization reaction by heating with polyphosphoric acid as well as by ch...

  3. Corrosion mechanisms for metal alloy waste forms: experiment and theory Level 4 Milestone M4FT-14LA0804024 Fuel Cycle Research & Development

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Taylor, Christopher D. [The Ohio State Univ., Columbus, OH (United States). Fontana Corrosion Center; Kim, Eunja [Univ. of Nevada, Las Vegas, NV (United States); Goff, George Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kolman, David Gary [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-07-31

    This document meets Level 4 Milestone: Corrosion mechanisms for metal alloy waste forms - experiment and theory. A multiphysics model is introduces that will provide the framework for the quantitative prediction of corrosion rates of metallic waste forms incorporating the fission product Tc. The model requires a knowledge of the properties of not only the metallic waste form, but also the passive oxide films that will be generated on the waste form, and the chemistry of the metal/oxide and oxide/environment interfaces. in collaboration with experimental work, the focus of this work is on obtaining these properties from fundamental atomistic models. herein we describe the overall multiphysics model, which is based on MacDonald's point-defect model for passivity. We then present the results of detailed electronic-structure calculations for the determination of the compatibility and properties of Tc when incorporated into intermetallic oxide phases. This work is relevant to the formation of multi-component oxides on metal surfaces that will incorporate Tc, and provide a kinetic barrier to corrosion (i.e. the release of Tc to the environment). Atomistic models that build upon the electronic structure calculations are then described using the modified embedded atom method to simulate metallic dissolution, and Buckingham potentials to perform classical molecular dynamics and statics simulations of the technetium (and, later, iron-technetium) oxide phases. Electrochemical methods were then applied to provide some benchmark information of the corrosion and electrochemical properties of Technetium metal. The results indicate that published information on Tc passivity is not complete and that further investigation is warranted.

  4. Remaining Sites Verification Package for the 128-B-3 Burn Pit Site, Waste Site Reclassification Form 2006-058

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-11-17

    The 128-B-3 waste site is a former burn and disposal site for the 100-B/C Area, located adjacent to the Columbia River. The 128-B-3 waste site has been remediated to meet the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results of sampling at upland areas of the site also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  5. 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.

  6. Demonstration of an approach to waste form qualification through simulation of liquid-fed ceramic melter process operations

    Energy Technology Data Exchange (ETDEWEB)

    Reimus, P.W.; Kuhn, W.L.; Peters, R.D.; Pulsipher, B.A.

    1986-07-01

    During fiscal year 1982, the US Department of Energy (DOE) assigned responsibility for managing civilian nuclear waste treatment programs in the United States to the Nuclear Waste Treatment Program (NWTP) at the Pacific Northwest Laboratory (PNL). One of the principal objectives of this program is to establish relationships between vitrification process control and glass quality. Users of the liquid-fed ceramic melter (LFCM) process will need such relationships in order to establish acceptance of vitrified high-level nuclear waste at a licensed federal repository without resorting to destructive examination of the canisters. The objective is to be able to supply a regulatory agency with an estimate of the composition, durability, and integrity of the glass in each waste glass canister produced from an LFCM process simply by examining the process data collected during the operation of the LFCM. The work described here will continue through FY-1987 and culminate in a final report on the ability to control and monitor an LFCM process through sampling and process control charting of the LFCM feed system.

  7. Remaining Sites Verification Package for the 128-F-2, 100-F Burning Pit Waste Site, Waste Site Reclassification Form 2008-031

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-12-01

    The 128-F-2 waste site consisted of multiple burn and debris filled pits located directly east of the 107-F Retention Basin and approximately 30.5 m east of the northeast corner of the 100-F Area perimeter road that runs along the riverbank. The burn pits were used for incinerating nonradioactive, combustible materials from 1945 to 1965. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  8. Usefulness of TAO model to predict and manage the transformation in soil of carbon and nitrogen forms from West-Africa urban solid wastes.

    Science.gov (United States)

    Kaboré, W T; Pansu, M; Hien, E; Houot, S; Zombré, N P; Masse, D

    2011-01-01

    The TAO model of Transformation of Added Organic materials (AOM) calibrated on AOMs and substrates of temperate areas was used to assess the transformations in soil of carbon and nitrogen forms of AOMs: raw materials, selected mixtures and composts from Ouagadougou urban wastes. AOMs were studied in terms of chemical and biochemical contents and for their C and N mineralization during incubations in a typical Ferric Lixisol of the sub-urban agriculture of Ouagadougou. The TAO model was used to predict the transformations of C (very labile, resistant and stable organic C) and N (very labile, resistant and stable organic N, produced and immobilized inorganic N) forms driven by AOM biochemical data. Without any change in calibration formulae, TAO predicted accurately the C transformations and inorganic N production of most of the tested AOMs, with a tendency to slightly overestimate C mineralization of previously well-composted materials and re-mineralization of immobilized N. Complementary adjustments using more complete data from laboratory experiments are suggested, but the model agrees with other data collected in the field and appears as a promising tool to optimise the management of urban wastes in the tropical area as well as for agro industrial organic fertilizers of the temperate zone. This application suggests ways to improve the management of urban wastes aiming to optimize agricultural yields, system sustainability and C sequestration in soil. Copyright © 2010 Elsevier Ltd. All rights reserved.

  9. Development of test acceptance standards for qualification of the glass-bonded zeolite waste form. Interim annual report, October 1995--September 1996

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, L.J.; Wronkiewicz, D.J.; Fortner, J.A.

    1997-09-01

    Glass-bonded zeolite is being developed at Argonne National Laboratory in the Electrometallurgical Treatment Program as a potential ceramic waste form for the disposition of radionuclides associated with the US Department of Energy`s (DOE`s) spent nuclear fuel conditioning activities. The utility of standard durability tests [e.g. Materials Characterization Center Test No. 1 (MCC-1), Product Consistency Test (PCT), and Vapor Hydration Test (VHT)] are being evaluated as an initial step in developing test methods that can be used in the process of qualifying this material for acceptance into the Civilian Radioactive Waste Management System. A broad range of potential repository conditions are being evaluated to determine the bounding parameters appropriate for the corrosion testing of the ceramic waste form, and its behavior under accelerated testing conditions. In this report we provide specific characterization information and discuss how the durability test results are affected by changes in pH, leachant composition, and sample surface area to leachant volume ratios. We investigate the release mechanisms and other physical and chemical parameters that are important for establishing acceptance parameters, including the development of appropriate test methodologies required to measure product consistency.

  10. Remaining Sites Verification Package for the 1607-F4 Sanitary Sewer System, Waste Site Reclassification Form 2004-131

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-12-03

    The 1607-F4 waste site is the former location of the sanitary sewer system that serviced the former 115-F Gas Recirculation Building. The system included a septic tank, drain field, and associated pipeline that were in use from 1944 to 1965. The 1607-F4 waste site received unknown amounts of sanitary sewage from the 115-F Gas Recirculation Building and may have potentially contained hazardous and radioactive contamination. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  11. Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank SX-105, Tank AN-103, And AZ-101/102) By Fluidized Bed Steam Reformation (FBSR)

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M.; Crawford, C. L.; Bannochie, C. J.; Burket, P. R.; Cozzi, A. D.; Daniel, W. E.; Hall, H. K.; Miller, D. H.; Missimer, D. M.; Nash, C. A.; Williams, M. F.

    2013-09-18

    Fluidized Bed Steam Reforming (FBSR) is a robust technology for the immobilization of a wide variety of radioactive wastes. Applications have been tested at the pilot scale for the high sodium, sulfate, halide, organic and nitrate wastes at the Hanford site, the Idaho National Laboratory (INL), and the Savannah River Site (SRS). Due to the moderate processing temperatures, halides, sulfates, and technetium are retained in mineral phases of the feldspathoid family (nepheline, sodalite, nosean, carnegieite, etc). The feldspathoid minerals bind the contaminants such as Tc-99 in cage (sodalite, nosean) or ring (nepheline) structures to surrounding aluminosilicate tetrahedra in the feldspathoid structures. The granular FBSR mineral waste form that is produced has a comparable durability to LAW glass based on the short term PCT testing in this study, the INL studies, SPFT and PUF testing from previous studies as given in the columns in Table 1-3 that represent the various durability tests. Monolithing of the granular product was shown to be feasible in a separate study. Macro-encapsulating the granular product provides a decrease in leaching compared to the FBSR granular product when the geopolymer is correctly formulated.

  12. Chelating, film-forming, and coagulating ability of the chitosan-glucan complex from Aspergillus niger industrial wastes.

    Science.gov (United States)

    Muzzarelli, R A; Tanfani, F; Scarpini, G

    1980-04-01

    Waste mycelia of Aspergillus niger from a citric acid production plant are simply treated with boiling 30-40% NaOH aqueous solutions for 4-6 hr to obtain the insoluble chitosan-glucan complex whose infrared, ESR, and x-ray diffraction spectra are reported. A number of transition- and post-transition-metal ions are chelated and collected by chitosan-glucan with higher yields than by animal chitosan. Immediate flocculation occurs upon mixing chitosan-glucan dispersions with alginate and polymolybdate solutions. Membranes are also obtained from chitosan-glucan dispersions in acetic acid or in chloral and dimethyl formamide mixtures.

  13. Remaining Sites Verification Package for the 1607-F3 Sanitary Sewer System, Waste Site Reclassification Form 2006-047

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-04-26

    The 1607-F3 waste site is the former location of the sanitary sewer system that supported the 182-F Pump Station, the 183-F Water Treatment Plant, and the 151-F Substation. The sanitary sewer system included a septic tank, drain field, and associated pipeline, all in use between 1944 and 1965. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  14. Hanford waste-form release and sediment interaction: A status report with rationale and recommendations for additional studies

    Energy Technology Data Exchange (ETDEWEB)

    Serne, R.J. (Pacific Northwest Lab., Richland, WA (USA)); Wood, M.I. (Westinghouse Hanford Co., Richland, WA (USA))

    1990-05-01

    This report documents the currently available geochemical data base for release and retardation for actual Hanford Site materials (wastes and/or sediments). The report also recommends specific laboratory tests and presents the rationale for the recommendations. The purpose of this document is threefold: to summarize currently available information, to provide a strategy for generating additional data, and to provide recommendations on specific data collection methods and tests matrices. This report outlines a data collection approach that relies on feedback from performance analyses to ascertain when adequate data have been collected. The data collection scheme emphasizes laboratory testing based on empiricism. 196 refs., 4 figs., 36 tabs.

  15. Remaining Sites Verification Package for 132-D-3, 1608-D Effluent Pumping Station, Waste Site Reclassification Form 2005-033

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Carlson

    2006-05-09

    Decommissioning and demolition of the 132-D-3 site, 1608-D Effluent Pumping Station was performed in 1986. Decommissioning included removal of equipment, water, and sludge for disposal as radioactive waste. The at- and below-grade structure was demolished to at least 1 m below grade and the resulting rubble buried in situ. The area was backfilled to grade with at least 1 m of clean fill and contoured to the surrounding terrain. Residual concentrations support future land uses that can be represented by a rural-residential scenario and pose no threat to groundwater or the Columbia River based on RESRAD modeling.

  16. Remaining Sites Verification Package for the 126-F-2, 183-F Clearwells, Waste Site Reclassification Form 2006-017

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Carlson

    2006-05-04

    The 126-F-2 site is the clearwell facility formerly used as part of the reactor cooling water treatment at the 183-F facility. During demolition operations in the 1970s, potentially contaminated debris was disposed in the eastern clearwell structure. The site has been remediated by removing all debris in the clearwell structure to the Environmental Restoration Disposal Facility. The results of radiological surveys and visual inspection of the remediated clearwell structure show neither residual contamination nor the potential for contaminant migration beyond the clearwell boundaries. The results of verification sampling at the remediation waste staging area demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  17. Mitigation of Hydrogen Gas Generation from the Reaction of Uranium Metal with Water in K Basin Sludge and Sludge Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-08

    corrosion rates in water alone and in simulated sludge were near or slightly below the metal-in-water rate while nitrate-free sludge/Aquaset II decreased rates by about a factor of 3. Addition of 1 M nitrate to simulated sludge decreased the corrosion rate by a factor of ~5 while 1 M nitrate in sludge/Aquaset II mixtures decreased the corrosion rate by ~2.5 compared with the nitrate-free analogues. Mixtures of simulated sludge with Aquaset II treated with 1 M nitrate had uranium corrosion rates about a factor of 8 to 10 lower than the water-only rate law. Nitrate was found to provide substantial hydrogen mitigation for immobilized simulant sludge waste forms containing Aquaset II or Aquaset II G clay. Hydrogen attenuation factors of 1000 or greater were determined at 60°C for sludge-clay mixtures at 1 M nitrate. Hydrogen mitigation for tests with PC and Aquaset II H (which contains PC) were inconclusive because of suspected failure to overcome induction times and fully enter into anoxic corrosion. Lessening of hydrogen attenuation at ~80°C and ~95°C for simulated sludge and Aquaset II was observed with attenuation factors around 100 to 200 at 1 M nitrate. Valuable additional information has been obtained on the ability of nitrate to attenuate hydrogen gas generation from solution, simulant K Basin sludge, and simulant sludge with immobilization agents. Details on characteristics of the associated reactions were also obtained. The present testing confirms prior work which indicates that nitrate is an effective agent to attenuate hydrogen from uranium metal corrosion in water and simulated K Basin sludge to show that it is also effective in potential candidate solidified K Basin waste forms for WIPP disposal. The hydrogen mitigation afforded by nitrate appears to be sufficient to meet the hydrogen generation limits for shipping various sludge waste streams based on uranium metal concentrations and assumed waste form loadings.

  18. DEMONSTRATION OF LEACHXS/ORCHESTRA CAPABILITIES BY SIMULATING CONSTITUENT RELEASE FROM A CEMENTITIOUS WASTE FORM IN A REINFORCED CONCRETE VAULT

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Meeussen, J.; Sloot, H.

    2010-03-31

    The objective of the work described in this report is to demonstrate the capabilities of the current version of LeachXS{trademark}/ORCHESTRA for simulating chemical behavior and constituent release processes in a range of applications that are relevant to the CBP. This report illustrates the use of LeachXS{trademark}/ORCHESTRA for the following applications: (1) Comparing model and experimental results for leaching tests for a range of cementitious materials including cement mortars, grout, stabilized waste, and concrete. The leaching test data includes liquid-solid partitioning as a function of pH and release rates based on laboratory column, monolith, and field testing. (2) Modeling chemical speciation of constituents in cementitious materials, including liquid-solid partitioning and release rates. (3) Evaluating uncertainty in model predictions based on uncertainty in underlying composition, thermodynamic, and transport characteristics. (4) Generating predominance diagrams to evaluate predicted chemical changes as a result of material aging using the example of exposure to atmospheric conditions. (5) Modeling coupled geochemical speciation and diffusion in a three layer system consisting of a layer of Saltstone, a concrete barrier, and a layer of soil in contact with air. The simulations show developing concentration fronts over a time period of 1000 years. (6) Modeling sulfate attack and cracking due to ettringite formation. A detailed example for this case is provided in a separate article by the authors (Sarkar et al. 2010). Finally, based on the computed results, the sensitive input parameters for this type of modeling are identified and discussed. The chemical speciation behavior of substances is calculated for a batch system and also in combination with transport and within a three layer system. This includes release from a barrier to the surrounding soil as a function of time. As input for the simulations, the physical and chemical properties of the

  19. Characteristics of potential repository wastes. Volume 3, Appendix 3A, ORIGEN2 decay tables for immobilized high-level waste; Appendix 3B, Interim high-level waste forms

    Energy Technology Data Exchange (ETDEWEB)

    1992-07-01

    This appendix presents the results of decay calculations using the ORIGEN2 code to determine the radiological properties of canisters of immobilized high-level waste as a function of decay time for decay times up to one million years. These calculations were made for the four HLW sites (West Valley Demonstration Project, Savannah River Site, Hanford Site, and Idaho National Engineering Laboratory) using the composition data discussed in the HLW section of this report. Calculated ({alpha},n) neutron production rates are also shown.

  20. Remaining Sites Verification Package for the 100-F-50 Stormwater Runoff Culvert, Waste Site Reclassification Form 2007-001

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-04-15

    The 100-F-50 waste site, part of the 100-FR-2 Operable Unit, is a steel stormwater runoff culvert that runs between two railroad grades in the south-central portion of the 100-F Area. The culvert exiting the west side of the railroad grade is mostly encased in concrete and surrounded by a concrete stormwater collection depression partially filled with soil and vegetation. The drain pipe exiting the east side of the railroad grade embankment is partially filled with soil and rocks. The 100-F-50 stormwater diversion culvert confirmatory sampling results support a reclassification of this site to no action. The current site conditions achieve the remedial action objectives and corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  1. Remaining Sites Verification Package for the 100-F-50 Stormwater Runoff Culvert, Waste Site Reclassification Form 2007-001

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-04-15

    The 100-F-50 waste site, part of the 100-FR-2 Operable Unit, is a steel stormwater runoff culvert that runs between two railroad grades in the south-central portion of the 100-F Area. The culvert exiting the west side of the railroad grade is mostly encased in concrete and surrounded by a concrete stormwater collection depression partially filled with soil and vegetation. The drain pipe exiting the east side of the railroad grade embankment is partially filled with soil and rocks. The 100-F-50 stormwater diversion culvert confirmatory sampling results support a reclassification of this site to no action. The current site conditions achieve the remedial action objectives and corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  2. Effects of heat treatment and formulation on the phase composition and chemical durability of the EBR-ll ceramic waste form.

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, W. E.; Dietz, N. L.; Janney, D. E.

    2006-01-31

    High-level radioactive waste salts generated during the electrometallurgical treatment of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor-II will be immobilized in a ceramic waste form (CWF). Tests are being conducted to evaluate the suitability of the CWF for disposal in the planned federal high-level radioactive waste repository at Yucca Mountain. In this report, the results of laboratory tests and analyses conducted to address product consistency and thermal stability issues called out in waste acceptance requirements are presented. The tests measure the impacts of (1) variations in the amounts of salt and binder glass used to make the CWF and (2) heat treatments on the phase composition and chemical durability of the waste form. A series of CWF materials was made to span the ranges of salt and glass contents that could be used during processing: between 5.0 and 15 mass% salt loaded into the zeolite (the nominal salt loading is 10.7%, and the process control range is 10.6 to 11.2 mass%), and between 20 and 30 mass% binder glass mixed with the salt-loaded zeolite (the nominal glass content is 25% and the process control range is 20 to 30 mass%). In another series of tests, samples of two CWF products made with the nominal salt and glass contents were reheated to measure the impact on the phase composition and durability: long-term heat treatments were conducted at 400 and 500 C for durations of 1 week, 4 weeks, 3 months, 6 months, and 1 year; short-term heat treatments were conducted at 600, 700, 800, and 850 C for durations of 4, 28, 52, and 100 hours. All of the CWF products that were made with different amounts of salt, zeolite, and glass and all of the heat-treated CWF samples were analyzed with powder X-ray diffraction to measure changes in phase compositions and subjected to 7-day product consistency tests to measure changes in the chemical durability. The salt loading had the greatest impact on phase composition and durability. A

  3. 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)

  4. Mine waste disposal leads to lower coral cover, reduced species richness and a predominance of simple coral growth forms on a fringing coral reef in Papua New Guinea.

    Science.gov (United States)

    Haywood, M D E; Dennis, D; Thomson, D P; Pillans, R D

    2016-04-01

    A large gold mine has been operating at the Lihir Island Group, Papua New Guinea since 1997. The mine disposes of waste rock in nearshore waters, impacting nearby coral communities. During 2010, 2012 we conducted photographic surveys at 73 sites within 40 km of the mine to document impacts of mining operations on the hard coral communities. Coral communities close to the mine (∼2 km to the north and south of the mine) were depaurperate, but surprisingly, coral cover and community composition beyond this range appeared to be relatively similar, suggesting that the mine impacts were limited spatially. In particular, we found mining operations have resulted in a significant decrease in coral cover (4.4% 1.48 km from the disposal site c.f. 66.9% 10.36 km from the disposal site), decreased species richness and a predominance of less complex growth forms within ∼2 km to the north and south of the mine waste disposal site. In contrast to the two 'snapshot' surveys of corals performed in 2010 and 2012, long term data (1999-2012) based on visual estimates of coral cover suggested that impacts on coral communities may have been more extensive than this. With global pressures on the world's coral reefs increasing, it is vital that local, direct anthropogenic pressures are reduced, in order to help offset the impacts of climate change, disease and predation.

  5. Chronic wasting disease and atypical forms of bovine spongiform encephalopathy and scrapie are not transmissible to mice expressing wild-type levels of human prion protein.

    Science.gov (United States)

    Wilson, Rona; Plinston, Chris; Hunter, Nora; Casalone, Cristina; Corona, Cristiano; Tagliavini, Fabrizio; Suardi, Silvia; Ruggerone, Margherita; Moda, Fabio; Graziano, Silvia; Sbriccoli, Marco; Cardone, Franco; Pocchiari, Maurizio; Ingrosso, Loredana; Baron, Thierry; Richt, Juergen; Andreoletti, Olivier; Simmons, Marion; Lockey, Richard; Manson, Jean C; Barron, Rona M

    2012-07-01

    The association between bovine spongiform encephalopathy (BSE) and variant Creutzfeldt-Jakob disease (vCJD) has demonstrated that cattle transmissible spongiform encephalopathies (TSEs) can pose a risk to human health and raises the possibility that other ruminant TSEs may be transmissible to humans. In recent years, several novel TSEs in sheep, cattle and deer have been described and the risk posed to humans by these agents is currently unknown. In this study, we inoculated two forms of atypical BSE (BASE and H-type BSE), a chronic wasting disease (CWD) isolate and seven isolates of atypical scrapie into gene-targeted transgenic (Tg) mice expressing the human prion protein (PrP). Upon challenge with these ruminant TSEs, gene-targeted Tg mice expressing human PrP did not show any signs of disease pathology. These data strongly suggest the presence of a substantial transmission barrier between these recently identified ruminant TSEs and humans.

  6. Processing of Radioactive Waste Solution with Zeolites (I) : Thermal-Transformations of Na, Cs and Sr Forms of Zeolites

    OpenAIRE

    Mimura, Hitoshi; KANNO, Takuji

    1980-01-01

    Thermal-transformations of several kinds of zeolites have been studied by means of differential thermal analysis (DTA), thermo-gravimetric analysis (TGA) and X-ray powder diffraction. Some synthetic zeolites (A, X, Y, mordenite), natural mordenite and clinoptilolite were used. Sodium forms of A and X zeolites recrystallized above 1000℃ to nepheline (NaAlSiO_4) , whereas the structure of zeolite Y, synthetic and natural mordenites, and clinoptilolite collapsed above around 900℃ and did not rec...

  7. Waste disposal package

    Science.gov (United States)

    Smith, M.J.

    1985-06-19

    This is a claim for a waste disposal package including an inner or primary canister for containing hazardous and/or radioactive wastes. The primary canister is encapsulated by an outer or secondary barrier formed of a porous ceramic material to control ingress of water to the canister and the release rate of wastes upon breach on the canister. 4 figs.

  8. Informative document waste plastics

    NARCIS (Netherlands)

    Nagelhout D; Sein AA; Duvoort GL

    1989-01-01

    This "Informative document waste plastics" forms part of a series of "informative documents waste materials". These documents are conducted by RIVM on the indstruction of the Directorate General for the Environment, Waste Materials Directorate, in behalf of the program of acti

  9. Hanford Site Secondary Waste Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H.

    2009-01-29

    Summary The U.S. Department of Energy (DOE) is making plans to dispose of 54 million gallons of radioactive tank wastes at the Hanford Site near Richland, Washington. The high-level wastes and low-activity wastes will be vitrified and placed in permanent disposal sites. Processing of the tank wastes will generate secondary wastes, including routine solid wastes and liquid process effluents, and these need to be processed and disposed of also. The Department of Energy Office of Waste Processing sponsored a meeting to develop a roadmap to outline the steps necessary to design the secondary waste forms. Representatives from DOE, the U.S. Environmental Protection Agency, the Washington State Department of Ecology, the Oregon Department of Energy, Nuclear Regulatory Commission, technical experts from the DOE national laboratories, academia, and private consultants convened in Richland, Washington, during the week of July 21-23, 2008, to participate in a workshop to identify the risks and uncertainties associated with the treatment and disposal of the secondary wastes and to develop a roadmap for addressing those risks and uncertainties. This report describes the results of the roadmap meeting in Richland. Processing of the tank wastes will generate secondary wastes, including routine solid wastes and liquid process effluents. The secondary waste roadmap workshop focused on the waste streams that contained the largest fractions of the 129I and 99Tc that the Integrated Disposal Facility risk assessment analyses were showing to have the largest contribution to the estimated IDF disposal impacts to groundwater. Thus, the roadmapping effort was to focus on the scrubber/off-gas treatment liquids with 99Tc to be sent to the Effluent Treatment Facility for treatment and solidification and the silver mordenite and carbon beds with the captured 129I to be packaged and sent to the IDF. At the highest level, the secondary waste roadmap includes elements addressing regulatory and

  10. Composting of urban solid wastes and agricultural wastes: the influence on the temperature and oxygen levels; Compostaje conjunto de la fraccion organica de residuos munipales (FORM) y residuos vegetales triturados: influencia sobre la temperatura y los niveles de oxigeno

    Energy Technology Data Exchange (ETDEWEB)

    Manzano Castro, S.; Perz Losada, C.; Soliva Torrento, M.

    1998-07-01

    The Superior School of Agriculture of Barcelona states in this paper that the compost of urban solid waste improvement is better if they are mixed with wastes of prune vegetables. This mixture improves the oxygen load and the temperatures that are reached. (Author) 26 refs.

  11. Microwave irradiated synthesis and characterization of 1, 4-phenylene bis-oxazoline form bis-(2-hydroxyethyl terephthalamide obtained by depolymerization of poly (ethylene terephthalate (PET bottle wastes

    Directory of Open Access Journals (Sweden)

    Yogesh S. Parab

    2012-04-01

    Full Text Available The aminolytic depolymerization of PET bottle waste with ethanolamine by conventional heating and microwave irradiation heating method was attempted with heterogeneous, recyclable acid catalysts such as beta zeolite (SiO2/ AlO2= 15 Na- form and montmorillonite KSF. The pure product bis-(2-hydroxyethyl terephthalamide (BHETA of aminolysis was obtained in good yield (85- 88%. The BHETA, thus obtained, was subjected to cyclization reaction by heating with polyphosphoric acid as well as by chlorination (using phosphoryl chloride, bromination (using red phosphorous and liquid bromine and nitration (conc. HNO3 + conc. H2SO4 followed by conventional and microwave irradiation heating in N,N- dimethyl formamide/ potassium carbonate solution. The product so obtained was 2, 2’-(1,4-phenylene–bis-(2-oxazoline (PBO, which has applications in polymer synthesis as a chain extender/ chain coupling agent or a cross linker. The productswere analyzed by FTIR, DSC, Mass and NMR (1H and 13C NMR.

  12. Remaining Sites Verification Package for the 100-F-31, 144-F Sanitary Sewer System, Waste Site Reclassification Form 2006-033

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-08-24

    The 100-F-31 waste site is a former septic system that supported the inhalation laboratories, also referred to as the 144-F Particle Exposure Laboratory (132-F-2 waste site), which housed animals exposed to particulate material. The 100-F-31 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  13. Informative document halogenated hydrocarbon-containing waste

    NARCIS (Netherlands)

    Verhagen H

    1992-01-01

    This "Informative document halogenated hydrocarbon-containing waste" forms part of a series of "Informative documents waste materials". These documents are conducted by RIVM on the instructions of the Directorate General for the Environment, Waste Materials Directorate, in behal

  14. Remaining Sites Verification Package for the 1607-F1 Sanitary Sewer System (124-F-1) and the 100-F-26:8 (1607-F1) Sanitary Sewer Pipelines Waste Sites, Waste Site Reclassification Form 2005-004

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-03-14

    The 100-F-26:8 waste site consisted of the underground pipelines that conveyed sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office to the 1607-F1 septic tank. The site has been remediated and presently exists as an open excavation. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  15. Remaining Sites Verification Package for the 1607-F1 Sanitary Sewer System (124-F-1) and the 100-F-26:8 (1607-F1) Sanitary Sewer Pipelines Waste Sites, Waste Site Reclassification Form 2005-004

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-03-14

    The 100-F-26:8 waste site consisted of the underground pipelines that conveyed sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office to the 1607-F1 septic tank. The site has been remediated and presently exists as an open excavation. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  16. Remaining Sites Verification Package for the 331 Life Sciences Laboratory Drain Field Septic System, Waste Site Reclassification Form 2008-020

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-10-16

    The 331 Life Sciences Laboratory Drain Field (LSLDF) septic system waste site consists of a diversion chamber, two septic tanks, a distribution box, and a drain field. This septic system was designed to receive sanitary waste water, from animal studies conducted in the 331-A and 331-B Buildings, for discharge into the soil column. However, field observations and testing suggest the 331 LSLDF septic system did not receive any discharges. In accordance with this evaluation, the confirmatory sampling results support a reclassification of the 331 LSLDF waste site to No Action. This site does not have a deep zone or other condition that would warrant an institutional control in accordance with the 300-FF-2 ROD under the industrial land use scenario.

  17. Waste Sites - Municipal Waste Operations

    Data.gov (United States)

    NSGIC Education | GIS Inventory — A Municipal Waste Operation is a DEP primary facility type related to the Waste Management Municipal Waste Program. The sub-facility types related to Municipal Waste...

  18. Remaining Sites Verification Package for the 100-D-24, 119-D Sample Building Drywell, Waste Site Reclassification Form 2006-004

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-09-19

    The 100-D-24 Sample Building Drywell waste site was a drywell that received drainage from a floor drain in the 119-D Sample Building. Confirmatory sampling was conducted on November 3, 2005. The waste site meets the remedial action objectives specified in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  19. One Step In-Situ Formed Magnetic Chitosan Nanoparticles as an Efficient Sorbent for Removal of Mercury Ions From Petrochemical Waste Water: Batch and Column Study

    Directory of Open Access Journals (Sweden)

    Rahbar

    2015-10-01

    Full Text Available Background In the recent years, mercury contamination has attracted great deal of attention due to its serious environmental threat. Objectives The main goal of this study was application of one-step synthesized magnetic (magnetite chitosan nanoparticles (MCNs in the removal of mercury ions from petrochemical waste water. Materials and Methods This study was performed in batch and column modes. Effects of various parameters such as pH, adsorbent dose, contact time, temperature and agitation speed for the removal of mercury ions by MCNs investigated in batch mode. Afterwards, optimum conditions were exploited in column mode. Different kinetic models were also studied. Results An effective Hg (II removal (99.8% was obtained at pH 6, with 50 mg of MCNs for an initial concentration of this ion in petrochemical waste water (5.63 mg L-1 and 10 minutes agitation of the solution. The adsorption kinetic data was well fitted to the pseudo-second-order model. Conclusions Experimental results showed that MCNs is an excellent sorbent for removal of mercury ions from petrochemical waste water. In addition, highly complex matrix of this waste does not affect the adsorption capability of MCNs.

  20. Characterization of the Italian glasses and their interaction with clay Task 3 Characterization of radioactive waste forms a series of final reports (1985-89) No 23

    Energy Technology Data Exchange (ETDEWEB)

    Cantale, C.; Castelli, S.; Donato, A.; Traverso, D.M. [ENEA, Casaccia (IT)

    1991-12-31

    The objective of this research work was the selection of a borosilicate glass composition suitable for the solidification of the HLW stream coming from the treatment of all the high-level wastes stored in Italy (MTR, Candu and Elk River) and the characterization of this glass with reference to the geological disposal. This research work was part of an Italian research project named `Ulisse project`, whose goal was the development and the demonstration of an integrated treatment of all the HLW stored in Italy, after their mixing (resulting waste: MCE waste). The main concept is to carry out a pre-treatment of the wastes, in order to concentrate the HLW fraction and to simplify the vitrification process, separating the most part of the inert salts. The research work concerning the separation process and pilot plant demonstration of the pre-treatment process were carried out in the framework of the CEC R and D programme (Contract No Fl1W-0011-lS). The laboratory studies concerning the vitrification of the resulting HLW streams and the vitrification demonstration in the Italian full-scale, inactive IVET plant complete the `Ulisse project`. Some glass compositions were prepared and preliminarily characterized. The glass named BAZ was finally selected. A complete characterization of this glass was carried out in order to evaluate its mechanical, physical and physico-chemical properties. The chemical durability was evaluated by the MCC-1 static leach test at 90{sup 0}C, using three different leachants and two surface-area to leachant-volume ratios. The same characterization programme was applied to the BAZ glass produced in the IVET plant during the plant vitrification demonstration programme. A comparison between the two glasses and a critical evaluation of their performances with respect to other nuclear waste glasses` durability was performed. 25 refs.; 46 figs.; 20 tabs.

  1. Properties of waste stillage from shochu distillery and waste water occurred sosei paper production process

    OpenAIRE

    山内, 正仁; 平田, 登基男; 前野, 祐二; 三原, めぐみ; 松藤, 康司

    1999-01-01

    As an effective utilization of waste stillage, which will be banned from being dumped into sea from the year of 2001, authors have been studied and succeeded to make the sosei paper by using waste stillage form shochu distillery. This research is tried to consider the property of waste stillage from shochu distillery ( sweet potato waste stillage and barley waste stillage) and the weight and property of waste water in compressing samples added some amount of old newspaper to waste stillage. F...

  2. Properties of waste stillage from shochu distillery and waste water occurred sosei paper production process

    OpenAIRE

    山内, 正仁; 平田, 登基男; 前野, 祐二; 三原, めぐみ; 松藤, 康司

    1999-01-01

    As an effective utilization of waste stillage, which will be banned from being dumped into sea from the year of 2001, authors have been studied and succeeded to make the sosei paper by using waste stillage form shochu distillery. This research is tried to consider the property of waste stillage from shochu distillery ( sweet potato waste stillage and barley waste stillage) and the weight and property of waste water in compressing samples added some amount of old newspaper to waste stillage. F...

  3. Waste to energy

    CERN Document Server

    Syngellakis, S

    2014-01-01

    Waste to Energy deals with the very topical subject of converting the calorific content of waste material into useful forms of energy. Topics included cover: Biochemical Processes; Conversions by Thermochemical Processes; Computational Fluid Dynamics Modelling; Combustion; Pyrolysis; Gasification; Biofuels; Management and Policies.

  4. Residential Waste

    DEFF Research Database (Denmark)

    Christensen, Thomas Højlund; Fruergaard, Thilde; Matsufuji, Y.

    2011-01-01

    are discussed in this chapter. Characterizing residential waste is faced with the problem that many residences already divert some waste away from the official collection systems, for example performing home composting of vegetable waste and garden waste, having their bundled newspaper picked up by the scouts...... twice a year or bringing their used furniture to the flea markets organized by charity clubs. Thus, much of the data available on residential waste represents collected waste and not necessarily all generated waste. The latter can only be characterized by careful studies directly at the source......Residential waste comes from residential areas with multi-family and single-family housing and includes four types of waste: household waste, garden waste, bulky waste and household hazardous waste. Typical unit generation rates, material composition, chemical composition and determining factors...

  5. Residential Waste

    DEFF Research Database (Denmark)

    Christensen, Thomas Højlund; Fruergaard, Thilde; Matsufuji, Y.

    2011-01-01

    Residential waste comes from residential areas with multi-family and single-family housing and includes four types of waste: household waste, garden waste, bulky waste and household hazardous waste. Typical unit generation rates, material composition, chemical composition and determining factors...... are discussed in this chapter. Characterizing residential waste is faced with the problem that many residences already divert some waste away from the official collection systems, for example performing home composting of vegetable waste and garden waste, having their bundled newspaper picked up by the scouts...... twice a year or bringing their used furniture to the flea markets organized by charity clubs. Thus, much of the data available on residential waste represents collected waste and not necessarily all generated waste. The latter can only be characterized by careful studies directly at the source...

  6. Melt Processed Single Phase Hollandite Waste Forms For Nuclear Waste Immobilization: Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, Kyle [Savannah River National Laboratory, Aiken, SC 29808 (United States); Marra, James [Savannah River National Laboratory, Aiken, SC 29808 (United States); Amoroso, Jake [Savannah River National Laboratory, Aiken, SC 29808 (United States); Conradson, Steven D. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Tang, Ming [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-09-23

    Cs is one of the more problematic fission product radionuclides to immobilize due to its high volatility at elevated temperatures, ability to form water soluble compounds, and its mobility in many host materials. The hollandite structure is a promising crystalline host for Cs immobilization and has been traditionally fabricated by solid state sintering methods. This study presents the structure and performance of Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al hollandite fabricated by melt processing. Melt processing is considered advantageous given that melters are currently in use for High Level Waste (HLW) vitrification in several countries. This work details the impact of Cr additions that were demonstrated to i) promote the formation of a Cs containing hollandite phase and ii) maintain the stability of the hollandite phase in reducing conditions anticipated for multiphase waste form processing.

  7. Remaining Sites Verification Package for the 100-F-26:13, 108-F Drain Pipelines, Waste Site Reclassification Form 2005-011

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-03-03

    The 100-F-26:13 waste site is the network of process sewer pipelines that received effluent from the 108-F Biological Laboratory and discharged it to the 188-F Ash Disposal Area (126-F-1 waste site). The pipelines included one 0.15-m (6-in.)-, two 0.2-m (8-in.)-, and one 0.31-m (12-in.)-diameter vitrified clay pipe segments encased in concrete. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  8. Remaining Sites Verification Package for the 1607-B2 Septic System and 100-B-14:2 Sanitary Sewer System, Waste Site Reclassification Form 2006-055

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-03-21

    The 1607-B2 waste site is a former septic system associated with various 100-B facilities, including the 105-B, 108-B, 115-B/C, and 185/190-B buildings. The site was evaluated based on confirmatory results for feeder lines within the 100-B-14:2 subsite and determined to require remediation. The 1607-B2 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  9. Remaining Sites Verification Package for the 116-F-8, 1904-F Outfall Structure and the 100-F-42, 1904-F Spillway, Waste Site Reclassification Form 2006-038

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-09-25

    The 116-F-8 waste site is the former 1904-F Outfall Structure used to discharge reactor cooling water effluent fro mthe 107-F Retention Basin to the Columbia River. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  10. Vitrification of hazardous and radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Bickford, D.F.; Schumacher, R.

    1995-12-31

    Vitrification offers many attractive waste stabilization options. Versatility of waste compositions, as well as the inherent durability of a glass waste form, have made vitrification the treatment of choice for high-level radioactive wastes. Adapting the technology to other hazardous and radioactive waste streams will provide an environmentally acceptable solution to many of the waste challenges that face the public today. This document reviews various types and technologies involved in vitrification.

  11. Chemical form distribution characters of heavy metals in waste PDP glass%废弃等离子显示屏玻璃中重金属化学形态分布

    Institute of Scientific and Technical Information of China (English)

    姜鹏飞; 陈海焱; 陈梦君; 蒋冬梅

    2012-01-01

    The chemical form of heavy metals from waste plasma display panel(PDP) glass was studied through BCR’s three stage(BCR3) extraction procedures and BCR’s four stage(BCR4) extraction procedures.The results show that the heavy metal extraction effect with BCR3 procedures was steadier than that with BCR4 procedures.There are significant differences in chemical form distributions of Zn,Cu and Ag in waste PDP glass.Both BCR3 and BCR4 think the chemical form of Ag in waste PDP glass exists mainly in residual forms and has little effect on the environment.The percentage of exchangeable and weak acid soluble form with BCR3 procedures or exchangeable metals and metals associated to carbonates form with BCR4 procedures of Cu are low,which has little risk for the environment.The estimates of the chemical form of Zn with BCR3 and BCR4 are inconformity,metals associated to oxides of Fe and Mn forms of Zn is the main form in waste PDP glass with BCR4 procedures,while the percentage of exchangeable and weak acid soluble form of Zn is the highest with BCR3 procedures,the harm of Zn from waste PDP glass is high.%通过BCR3步和BCR4步连续提取法,对等离子体显示屏玻璃中重金属的化学形态进行研究。结果表明,BCR3步连续提取法对废弃等离子体显示屏玻璃中重金属的提取比BCR4步连续提取法更稳定。废弃等离子体显示屏玻璃中Zn、Cu和Ag化学形态分布存在差异。BCR3步和BCR4步连续提取法均认为Ag以残渣态为主要赋存形态,对环境影响比较小;Cu的可交换与弱酸可溶态或交换态和碳酸盐结合态含量较少,潜在的环境风险也比较小;BCR3步和BCR4步连续提取法对Zn的形态分布结果评价不一致,BCR3步连续提取法表明Zn主要以可交换与弱酸可溶态为主,对环境影响较大,而BCR4步连续提取法结果认为Zn是以铁锰氧化态为主要存在形态。

  12. WASTE PACKAGE TRANSPORTER DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    D.C. Weddle; R. Novotny; J. Cron

    1998-09-23

    The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''.

  13. Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

    Energy Technology Data Exchange (ETDEWEB)

    Fachinger, Johannes; Muller, Walter [FNAG ZU Hanau, Hanau (Germany); Marsat, Eric [FNAG SAS Le Pont de Claix (France); Grosse, Karl-Heinz; Seemann, Richard [ALD Hanau (Germany); Scales, Charlie; Easton, Michael Mark [NNL, Workington (United Kingdom); Anthony Banford [NNL, Warrington (United Kingdom); University of Manchester, Manchester (United Kingdom)

    2013-07-01

    Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or

  14. Idaho Waste Vitrification Facilities Project Vitrified Waste Interim Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bonnema, Bruce Edward

    2001-09-01

    This feasibility study report presents a draft design of the Vitrified Waste Interim Storage Facility (VWISF), which is one of three subprojects of the Idaho Waste Vitrification Facilities (IWVF) project. The primary goal of the IWVF project is to design and construct a treatment process system that will vitrify the sodium-bearing waste (SBW) to a final waste form. The project will consist of three subprojects that include the Waste Collection Tanks Facility, the Waste Vitrification Facility (WVF), and the VWISF. The Waste Collection Tanks Facility will provide for waste collection, feed mixing, and surge storage for SBW and newly generated liquid waste from ongoing operations at the Idaho Nuclear Technology and Engineering Center. The WVF will contain the vitrification process that will mix the waste with glass-forming chemicals or frit and turn the waste into glass. The VWISF will provide a shielded storage facility for the glass until the waste can be disposed at either the Waste Isolation Pilot Plant as mixed transuranic waste or at the future national geological repository as high-level waste glass, pending the outcome of a Waste Incidental to Reprocessing determination, which is currently in progress. A secondary goal is to provide a facility that can be easily modified later to accommodate storage of the vitrified high-level waste calcine. The objective of this study was to determine the feasibility of the VWISF, which would be constructed in compliance with applicable federal, state, and local laws. This project supports the Department of Energy’s Environmental Management missions of safely storing and treating radioactive wastes as well as meeting Federal Facility Compliance commitments made to the State of Idaho.

  15. Radioactive Waste Streams: Waste Classification for Disposal

    Science.gov (United States)

    2006-12-13

    acidity with caustic soda or sodium nitrate to condition it for storage in the carbon-steel tanks. (The neutralization reaction formed a...waste ranges between from 47 to 147 curies/cubic-meter based on the Waste Isolation Pilot Plant inventory. The vitrified high-level waste processed by...Facility St T Assembly MTHM 1. Arkansas Nuclear One AK P 1,517 666.7 46. Shearon Harris Nuclear Power Plant NC P 3,814 964.5 I 552 241.4 47. Cooper

  16. Remaining Sites Verification Package for the 100-F-26:15 Miscellaneous Pipelines Associated with the 132-F-6, 1608-F Waste Water Pumping Station, Waste Site Reclassification Form 2007-031

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-03-18

    The 100-F-26:15 waste site consisted of the remnant portions of underground process effluent and floor drain pipelines that originated at the 105-F Reactor. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  17. Hazardous Waste

    Science.gov (United States)

    ... you throw these substances away, they become hazardous waste. Some hazardous wastes come from products in our homes. Our garbage can include such hazardous wastes as old batteries, bug spray cans and paint ...

  18. 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)

  19. Nevada Test Site Waste Acceptance Criteria, December 2000

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-12-01

    This document establishes the US Department of Energy, Nevada Operations Office waste acceptance criteria. The waste acceptance criteria provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive waste and mixed waste for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the Nevada Test Site Area 3 and Area 5 Radioactive Waste Management Sites for storage or disposal.

  20. Radioactive waste management in Austria

    OpenAIRE

    Neubauer Josef

    2004-01-01

    At the Austrian Research Centers Seibersdorf, there are several facilities in stalled for treatment of waste of low and intermediate radioactivity level (radwaste). A separate company within Centers, Nuclear Engineering Seibersdorf, has been formed recently, acting as a centralized facility for treatment, conditioning and storing of such waste within the country. The relevant treatment technology is applied depending on the waste category. In total about 6900 m3 of solid waste of low and inte...

  1. nfluences of ammonium-nitrate, food waste compost and bacterial fertilizer on soluble soil nitrogen forms and on the growth of carrot (Daucus Carota L.

    Directory of Open Access Journals (Sweden)

    Andrea Balla Kovács

    2014-04-01

    Full Text Available This paper reports a greenhouse study to compare the effects of food waste compost, bacterial fertilizer and their combination with the effect of mineral fertilizer on yield of carrot and the available nutrient content of soils. The study was conducted on calcareous chernozem and acidic sandy soils and consisted of 8 treatments in a randomized complete block design with four replications. The NH4NO3 resulted in reduced growing of carrot plant in sandy soil, and the treatment effect of mineral fertilizer was not observed significantly in chernozem soil. Sandy soil showed higher response of growth of carrot to food waste compost fertilization than chernozem soil. Sole application of EM-1 bacterial fertilizer did not have marked effect on yield parameters and sizes of roots. When EM-1 bacterial fertilizer was applied together with ammonium-nitrate or with compost in chernozem soil, the weights of roots and the sizes of roots in some cases became higher compared to the values of appropriate treatments without inoculation. In sandy soil the diameter of roots slightly increased when EM-1 bacterial fertilizer was applied with ammonium-nitrate and with ammonium-nitrate+compost combination compared to appropriate treatment without inoculation. In chernozem soil the maximum weights and sizes of roots were achieved with the combined treatment of ammonium-nitrate+compost+EM-1 bacterial fertilizer and in sandy soil with compost treatment. Our results of soluble nitrogen content of soils are in good agreement with yield parameters of carrot. Results suggest that food waste compost could be a good substitute for mineral fertilizer application in carrot production mainly in sandy soil. EM-1 bacterial fertilizer did not cause marked effect on yield and yield parameters of carrot plant, but its combination with other fertilizers promises a little bit higher yield or plant available nutrient in the soil. These effects do not clear exactly, so further studies are

  2. Remaining Sites Verification Package for the 100-F-33, 146-F Aquatic Biology Fish Ponds, Waste Site Reclassification Form 2006-021

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-08-25

    The 100-F-33, 146-F Aquatice Biology Fish Ponds waste site was an area with six small rectangular ponds and one large circular pond used to conduct tests on fish using various mixtures of river and reactor effluent water. The current site conditions achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification and applicable confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  3. Remaining Sites Verification Package for the 100-B-22:1 Pipelines and Associated Soils, Waste Site Reclassification Form 2005-042

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-09-12

    The 100-B-22:1 pipelines and associated soils were part of the 100-B Area water treatment facilities. The 100-B-22:1 waste site is limited to those pipelines that interconnected the 185-B Filter House, the 126-B-2 Clearwells, the 185-B Deaeration Plant, and the 190-B Process Pumphouse. None of the 100-B-22:1 pipelines carried environmentally significant contamination. In accordance with the historical information and field observations of this evaluation, the results support a reclassification of this site to No Action required to meet future rural-residential uses and be protective of groundwater and the Columbia River.

  4. Remaining Sites Verification Package for the 116-F-8, 1904-F Outfall Structure and the 100-F-42, 1904-F Spillway, Waste Site Reclassification Form 2006-045

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-09-26

    The 100-F-42 waste site is the portion of the former emergency overflow spillway for the 1904-F Outfall Structure formerly existing above the ordinary high water mark of the Columbia River. The spillway consisted of a concrete flume designed to discharge effluent from the 107-F Retention Basin in the event that flows could not be completely discharged via the river outfall pipelines. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  5. Remaining Sites Verification Package for the 126-B-3, 184-B Coal Pit Dumping Area, Waste Site Reclassification Form 2005-028

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-08-07

    The 126-B-3 waste site is the former coal storage pit for the 184-B Powerhouse. During demolition operations in the 1970s, the site was used for disposal of demolition debris from 100-B/C Area facilities. The site has been remediated by removing debris and contaminated soils. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  6. Remaining Sites Verification Package for the 141-C Large Animal Barn and Biology Laboratory (Hog Barn), Waste Site Reclassification Form 2006-027

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Carlson

    2006-05-24

    The 141-C waste site is a former large animal barn and biology laboratory within the 100-F Area experimental animal farm. Strontium-90, arsenic, and multiple polycyclic aromatic hydrocarbons were detected within residual demolition debris at concentrations exceeding cleanup criteria. The site has been remediated by removing approximately 900 bank cubic meters of soil and debris within the former building footprint to the Environmental Restoration Disposal Facility. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  7. Remaining Sites Verification Package for the 1607-F1 Sanitary Sewer System (124-F-1) and the 100-F-26:8 (1607-F1) Sanitary Sewer Pipelines Waste Sites, Waste Site Reclassification Form 2004-130

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-03-14

    The 1607-F1 Sanitary Sewer System (124-F-1), consisted of a septic tank, drain field, and associated pipelines that received sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office via the 100-F-26:8 pipelines. The septic tank required remedial action based on confirmatory sampling. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  8. Remaining Sites Verification Package for the 1607-F1 Sanitary Sewer System (124-F-1) and the 100-F-26:8 (1607-F1) Sanitary Sewer Pipelines Waste Sites, Waste Site Reclassification Form 2004-130

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-03-14

    The 1607-F1 Sanitary Sewer System (124-F-1), consisted of a septic tank, drain field, and associated pipelines that received sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office via the 100-F-26:8 pipelines. The septic tank required remedial action based on confirmatory sampling. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  9. CLAB Transuranic Waste Spreadsheets

    Energy Technology Data Exchange (ETDEWEB)

    Leyba, J.D.

    2000-08-11

    The Building 772-F Far-Field Transuranic (TRU) Waste Counting System is used to measure the radionuclide content of waste packages produced at the Central Laboratory Facilities (CLAB). Data from the instrument are entered into one of two Excel spreadsheets. The waste stream associated with the waste package determines which spreadsheet is actually used. The spreadsheets calculate the necessary information required for completion of the Transuranic Waste Characterization Form (OSR 29-90) and the Radioactive Solid Waste Burial Ground Record (OSR 7-375 or OSR 7-375A). In addition, the spreadsheets calculate the associated Low Level Waste (LLW) stream information that potentially could be useful if the waste container is ever downgraded from TRU to LLW. The spreadsheets also have the capability to sum activities from source material added to a waste container after assay. A validation data set for each spreadsheet along with the appropriate results are also presented in this report for spreadsheet verification prior to each use.

  10. Remaining Sites Verification Package for the 100-B-18, 184-B Powerhouse Debris Pile, Waste Site Reclassification Form 2007-020

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-11-30

    The 100-B-18 Powerhouse Debris Pile contained miscellaneous demolition waste from the decommissioning activities of the 184-B Powerhouse. The debris covered an area roughly 15 m by 30 m and included materials such as concrete blocks, mixed aggregate/concrete slabs, stone rubble, asphalt rubble, traces of tar/coal, broken fluorescent lights, brick chimney remnants, and rubber hoses. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  11. Remaining Sites Verification Package for the 1607-F5 Sanitary Sewer System (124-F-5), Waste Site Reclassification Form 2006-043

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-09-14

    The 1607-F5 waste site is a former septic tank, tile field, and associated pipeline located within the 100-FR-1 Operable Unit that received sewage from the former 181-F Pumphouse. Lead, gamma-chlordane, and heptachlor epoxide were identified within or around the septic system at concentrations exceeding the direct exposure cleanup criteria. Multiple metal and pesticide constituents were also identified as exceeding the groundwater and river protection cleanup criteria. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  12. Remaining Sites Verification Package for the 100-F-44:2, Discovery Pipeline Near 108-F Building, Waste Site Reclassification Form 2007-006

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-05-30

    The 100-F-44:2 waste site is a steel pipeline that was discovered in a junction box during confirmatory sampling of the 100-F-26:4 pipeline from December 2004 through January 2005. The 100-F-44:2 pipeline feeds into the 100-F-26:4 subsite vitrified clay pipe (VCP) process sewer pipeline from the 108-F Biology Laboratory at the junction box. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  13. Remaining Sites Verification Package for the 100-F-26:12, 1.8-m (72-in.) Main Process Sewer Pipeline, Waste Site Reclassification Form 2007-034

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-04-29

    The 100-F-26:12 waste site was an approximately 308-m-long, 1.8-m-diameter east-west-trending reinforced concrete pipe that joined the North Process Sewer Pipelines (100-F-26:1) and the South Process Pipelines (100-F-26:4) with the 1.8-m reactor cooling water effluent pipeline (100-F-19). In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  14. Remaining Sites Verification Package for the 100-F-52, 146-FR Radioecology and Aquatic Biology Laboratory Soil, Waste Site Reclassification Form 2008-022

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-06-27

    The 100-F-52 waste site consisted of the soil under and around the former 146-FR Radioecology and Aquatic Biology Laboratory. The laboratory was used for studies of the effects of pre-reactor and post-reactor process water on fish eggs, young fish, and other small river creatures of interest. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  15. Remaining Sites Verification Package for the 100-F-44:2, Discovery Pipeline Near 108-F Building, Waste Site Reclassification Form 2007-006

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-05-30

    The 100-F-44:2 waste site is a steel pipeline that was discovered in a junction box during confirmatory sampling of the 100-F-26:4 pipeline from December 2004 through January 2005. The 100-F-44:2 pipeline feeds into the 100-F-26:4 subsite vitrified clay pipe (VCP) process sewer pipeline from the 108-F Biology Laboratory at the junction box. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  16. Remaining Sites Verification Package for the 100-C-9:1 Main Process Sewer Collection Line, Waste Site Reclassification Form 2004-012

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-06-11

    The 100-C-9:1 main process sewer pipeline, also known as the twin box culvert, was a dual reinforced process sewer that collected process effluent from the 183-C and 190-C water treatment facilities, discharging at the 132-C-2 Outfall. For remedial action purposes, the 100-C-9:1 waste site was subdivided into northern and southern sections. The 100-C-9:1 subsite has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  17. ZERO WASTE

    OpenAIRE

    Upadhyaya, Luv

    2013-01-01

    The aim of the thesis was to develop a clear vision on better waste management system. The thesis introduced the sustainable waste management along with innovation. The aim of the research was to find out the types of waste being introduced to environment, their consequence on human beings and surroundings, best policies, principles and practices to minimize the effect of the waste to lowest. The study was based on literature. The thesis includes the introduction of types of waste, clarifi...

  18. DOE Waste Treatability Group Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Kirkpatrick, T.D.

    1995-01-01

    This guidance presents a method and definitions for aggregating U.S. Department of Energy (DOE) waste into streams and treatability groups based on characteristic parameters that influence waste management technology needs. Adaptable to all DOE waste types (i.e., radioactive waste, hazardous waste, mixed waste, sanitary waste), the guidance establishes categories and definitions that reflect variations within the radiological, matrix (e.g., bulk physical/chemical form), and regulated contaminant characteristics of DOE waste. Beginning at the waste container level, the guidance presents a logical approach to implementing the characteristic parameter categories as part of the basis for defining waste streams and as the sole basis for assigning streams to treatability groups. Implementation of this guidance at each DOE site will facilitate the development of technically defined, site-specific waste stream data sets to support waste management planning and reporting activities. Consistent implementation at all of the sites will enable aggregation of the site-specific waste stream data sets into comparable national data sets to support these activities at a DOE complex-wide level.

  19. Behaviour of IPG waste forms bearing BaSO{sub 4} as the dominant sludge constituent generated from the treatment of water used for cooling the stricken power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Amamoto, Ippei; Kobayashi, Hidekazu; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki [Japan Atomic Energy Agency, Tokai-mura, Ibaraki Pref. (Japan); Kitamura, Naoto; Takebe, Hiromichi [Ehime University, Matsuyama, Ehime Pref. (Japan); Mitamura, Naoki; Tsuzuki, Tatsuya [Central Glass Co.,Ltd., Matsusaka, Mie Pref. (Japan)

    2013-07-01

    The great amount of water used for cooling the stricken power reactors at Fukushima Dai-ichi following the earthquake and tsunami of 11 March 2011 has resulted in accumulation of cooling water so-called the remaining water in some buildings. As the cooling water is subsequently contaminated by fission products (FPs) and some other radioactive substances, it is necessary to decontaminate this 'cooling water' to reduce the volume of liquid radioactive waste and to reuse it again for cooling the affected reactors. Some methods are applied to remove the radioactive substances from the cooling water. However, after treatments of water, there arises a secondary radioactive waste, the sludge. Steps are now taken to immobilize this sludge. In this paper, BaSO{sub 4}, as one of main constituents of the sludge, was chosen as an immobilizing target substance. The appropriate manufacturing condition of glass waste form for loading the sludge (BaSO{sub 4}) was studied and the chemical durability was evaluated by measuring the dissolution rate. For this experiment, the iron phosphate glass (IPG), which is known to possess a large loading capacity for a variety of chemical substances, was employed as the glass medium. Based on experimental results, it is evident that BaSO{sub 4} can be loaded into the IPG medium when it possesses the appropriate composition and melting temperature. During loading, BaSO{sub 4} converted into BaO, acting as a network modifier, which leads to enhanced stability of IPG. (authors)

  20. Storing Waste in Ceramic

    Energy Technology Data Exchange (ETDEWEB)

    Bourcier, W L; Sickafus, K

    2004-07-20

    Not all the nuclear waste destined for Yucca Mountain is in the form of spent fuel. Some of it will be radioactive waste generated from the production of nuclear weapons. This so-called defense waste exists mainly as corrosive liquids and sludge in underground tanks. An essential task of the U.S. high-level radioactive waste program is to process these defense wastes into a solid material--called a waste form. An ideal waste form would be extremely durable and unreactive with other repository materials. It would be simple to fabricate remotely so that it could be safely transported to a repository for permanent storage. What's more, the material should be able to tolerate exposure to intense radiation without degradation. And to minimize waste volume, the material must be able to contain high concentrations of radionuclides. The material most likely to be used for immobilization of radioactive waste is glass. Glasses are produced by rapid cooling of high-temperature liquids such that the liquid-like non-periodic structure is preserved at lower temperatures. This rapid cooling does not allow enough time for thermodynamically stable crystalline phases (mineral species) to form. In spite of their thermodynamic instability, glasses can persist for millions of years. An alternate to glass is a ceramic waste form--an assemblage of mineral-like crystalline solids that incorporate radionuclides into their structures. The crystalline phases are thermodynamically stable at the temperature of their synthesis; ceramics therefore tend to be more durable than glasses. Ceramic waste forms are fabricated at temperatures below their melting points and so avoid the danger of handling molten radioactive liquid--a danger that exists with incorporation of waste in glasses. The waste form provides a repository's first line of defense against release of radionuclides. It, along with the canister, is the barrier in the repository over which we have the most control. When a waste

  1. The effects of concentrate added to pineapple (Ananas comosus Linn. Mer.) waste silage in differing ratios to form complete diets, on digestion, excretion of urinary purine derivatives and blood metabolites in growing, male, Thai swamp buffaloes.

    Science.gov (United States)

    Jetana, T; Suthikrai, W; Usawang, S; Vongpipatana, C; Sophon, S; Liang, J B

    2009-04-01

    Four, male, growing Thai swamp buffaloes (197 +/- 5.3 kg and all 1 year old) were used to evaluate the effects of concentrate added to pineapple waste silage in differing ratios, to form a complete diet, studying in vivo digestion, the rate of passage, microbial protein synthesis and blood metabolites. Animals were fed ad libitum with 4 diets, using four combinations of pineapple waste silage (P) and concentrate (C), in the proportions (on a dry matter basis) of 0.8:0.2 (P80:C20), 0.6:0.4 (P60:C40), 0.4:0.6 (P40:C60) and 0.2:0.8 (P20:C80). The results showed that the intakes of dry matter (DM), organic matter (OM), nitrogen (N), the N-balance, urinary purine derivatives (PD) excretion, the ratios of allantoin to creatinine (CR), PD to CR, the plasma urea-N (PUN) and insulin increased in the animals, but the intake of neutral detergent fiber (NDF), the coefficient of whole tract, apparent digestibility of NDF, the transit time (TT) and the mean retention time (TMRT) decreased, when the proportion of concentrate in the diet increased. This study indicated that the proportion of P40:C60 in the diet produced the best efficiency of urinary PD excretion (mmol) per digestible OM intake (kg DOMI).

  2. NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA

    Energy Technology Data Exchange (ETDEWEB)

    U.S. DEPARTMENT OF ENERGY, NATIONAL NUCLEAR SECURITY ADMINISTRATION, NEVADA SITE OFFICE

    2005-07-01

    This document establishes the U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive and mixed waste for disposal. Mixed waste generated within the State of Nevada by NNSA/NSO activities is accepted for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the Nevada Test Site Area 3 and Area 5 Radioactive Waste Management Site for storage or disposal.

  3. NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA

    Energy Technology Data Exchange (ETDEWEB)

    U.S. DEPARTMENT OF ENERGY, NATIONAL NUCLEAR SECURITY ADMINISTRATION, NEVADA SITE OFFICE

    2005-07-01

    This document establishes the U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive and mixed waste for disposal. Mixed waste generated within the State of Nevada by NNSA/NSO activities is accepted for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the Nevada Test Site Area 3 and Area 5 Radioactive Waste Management Site for storage or disposal.

  4. Waste management

    DEFF Research Database (Denmark)

    2000-01-01

    The case study deals with public accountability issues connected to household waste management in the municipality of Copenhagen, Denmark.......The case study deals with public accountability issues connected to household waste management in the municipality of Copenhagen, Denmark....

  5. Waste management

    DEFF Research Database (Denmark)

    2000-01-01

    The case study deals with public accountability issues connected to household waste management in the municipality of Copenhagen, Denmark.......The case study deals with public accountability issues connected to household waste management in the municipality of Copenhagen, Denmark....

  6. Wasting and stunting - similarities and differences

    DEFF Research Database (Denmark)

    Briend, André; Khara, Tanya; Dolan, Carmel

    2015-01-01

    Wasting and stunting are often presented as two separate forms of malnutrition requiring different interventions for prevention and/or treatment. These two forms of malnutrition, however, are closely related and often occur together in the same populations and often in the same children. Wasting ...... in the absence of wasting and even in overweight children. Thus, food supplementation should be used with caution in populations where stunting is not associated with wasting and low fat stores....

  7. Radioactive Waste.

    Science.gov (United States)

    Blaylock, B. G.

    1978-01-01

    Presents a literature review of radioactive waste disposal, covering publications of 1976-77. Some of the studies included are: (1) high-level and long-lived wastes, and (2) release and burial of low-level wastes. A list of 42 references is also presented. (HM)

  8. Teaching Form as Form

    DEFF Research Database (Denmark)

    Keiding, Tina Bering

    2012-01-01

    understanding of form per se, or, to use an expression from this text, of form as form. This challenge can be reduced to one question: how can design teaching support students in achieving not only the ability to recognize and describe different form-related concepts in existing design (i.e. analytical...... means that form serves both as the connective value and as the concept for reflection. In other words, form is observed as form, not anything else. The didactical challenge of teaching form as form is accentuated by students’ everyday-based pre-orientation towards function at the expense of form....... In general, students enter design education as far more skilled observers with regards to function than form. They are, in other words, predisposed to observe objects asking ‘what is?’, rather than ‘how is?’. This habit has not only cognitive implications. It is closely intertwined with a rudimentary...

  9. Teaching Form as Form

    DEFF Research Database (Denmark)

    Keiding, Tina Bering

    2012-01-01

    understanding of form per se, or, to use an expression from this text, of form as form. This challenge can be reduced to one question: how can design teaching support students in achieving not only the ability to recognize and describe different form-related concepts in existing design (i.e. analytical...... means that form serves both as the connective value and as the concept for reflection. In other words, form is observed as form, not anything else. The didactical challenge of teaching form as form is accentuated by students’ everyday-based pre-orientation towards function at the expense of form...... vocabulary of form. Even in cases in which teaching uses terms and phrases from everyday life (for instance, ‘intersection’), the meaning of the word cannot necessarily be transmitted directly from an ordinary vocabulary into a design context. And it is clearly a common issue for the contributions...

  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. Efficacy of a Solution-Based Approach for Making Sodalite Waste Forms for an Oxide Reduction Salt Utilized in the Reprocessing of Used Uranium Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyas, Josef; Burns, Carolyn A.

    2015-04-01

    This paper describes various approaches for making sodalite with a LiCl-Li2O oxide reduction salt used to recover uranium from used oxide fuel. The approaches include sol-gel and solution-based synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3-SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2O and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt.

  12. Efficacy of a solution-based approach for making sodalite waste forms for an oxide reduction salt utilized in the reprocessing of used uranium oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J., E-mail: brian.riley@pnnl.gov [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Pierce, David A. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Frank, Steven M. [Idaho National Laboratory, Idaho Falls, ID 83402 (United States); Matyáš, Josef; Burns, Carolyne A. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

    2015-04-15

    This paper describes the various approaches evaluated for making solution-derived sodalite with a LiCl–Li{sub 2}O oxide reduction salt selected to dissolve used uranium oxide fuel so the uranium can be recovered and recycled. The approaches include modified sol–gel and solution-based synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na{sub 2}O–B{sub 2}O{sub 3}–SiO{sub 2} glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na{sup +} and Cl{sup −} to form halite in solution and Li{sub 2}O and SiO{sub 2} to form lithium silicates (e.g., Li{sub 2}SiO{sub 3} or Li{sub 2}Si{sub 2}O{sub 5}) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions in the crystalline product (∼92 mass%) and low porosities using a solution-based approach and this LiCl–Li{sub 2}O salt but that the incorporation of Li into the sodalite is low.

  13. Industrial Waste

    DEFF Research Database (Denmark)

    Christensen, Thomas Højlund

    2011-01-01

    of the system industry has to inform at the planning stage and afterwards in yearly reports on their waste arising and how the waste is managed. If available such information is very helpful in obtaining information about that specific industry. However, in many countries there is very little information......Industrial waste is waste from industrial production and manufacturing. Industry covers many industrial sectors and within each sector large variations are found in terms of which raw materials are used, which production technology is used and which products are produced. Available data on unit...... generation rates and material composition as well as determining factors are discussed in this chapter. Characterizing industrial waste is faced with the problem that often only a part of the waste is handled in the municipal waste system, where information is easily accessible. In addition part...

  14. Aluminum phosphate ceramics for waste storage

    Science.gov (United States)

    Wagh, Arun; Maloney, Martin D

    2014-06-03

    The present disclosure describes solid waste forms and methods of processing waste. In one particular implementation, the invention provides a method of processing waste that may be particularly suitable for processing hazardous waste. In this method, a waste component is combined with an aluminum oxide and an acidic phosphate component in a slurry. A molar ratio of aluminum to phosphorus in the slurry is greater than one. Water in the slurry may be evaporated while mixing the slurry at a temperature of about 140-200.degree. C. The mixed slurry may be allowed to cure into a solid waste form. This solid waste form includes an anhydrous aluminum phosphate with at least a residual portion of the waste component bound therein.

  15. Abstraction’s Ecologies: Post-Industrialization, Waste and the Commodity Form in Prunella Clough’s Paintings of the 1980s and 1990s

    Directory of Open Access Journals (Sweden)

    Catherine Spencer

    2015-11-01

    Full Text Available This article’s aims are twofold: firstly, it argues that Prunella Clough’s engagement with consumer items in her paintings of the 1980s and 1990s constitute a sustained engagement with the fluctuating nature of the commodity form, moving beyond the established critical narrative whereby these works are understood as simply redeeming “everyday” materials. Secondly, in order to do this, it proposes new artistic frameworks for Clough’s work, moving away from her early association with Neo-Romanticism to foreground her relationship with Pop and Minimalism, and with Post-Conceptual painting. Clough’s late works, it finds, powerfully condense histories of industrial production and painting in Britain.

  16. Preparation of Y2Ti2O7 pyrochlore glass-ceramics as potential waste forms for actinides: The effects of processing conditions

    Science.gov (United States)

    Kong, Linggen; Zhang, Yingjie; Karatchevtseva, Inna

    2017-10-01

    Glass-Y2Ti2O7 pyrochlore was fabricated by sintering the mixture of glass precursor powder and (YTi)-composite which was prepared by a soft chemistry route. X-ray diffraction and Raman spectroscopy confirmed that the phase pure pyrochlore was crystallized in-situ in amorphous glass matrix at 1200 °C with dwelling time over 1 h. Pyrochlore was formed in glass matrix with cooling rate ∼0.1-40 °C/min, and independent of addition of alkali/alkaline earth fluorides. Glass matrix was able to accommodate/dissolve small amounts of impurities and the mean pyrochlore particle size was between 1 and 2 μm in glass observed by scanning electron microscopy.

  17. Nuclear Waste and Ethics

    Energy Technology Data Exchange (ETDEWEB)

    Damveld, Herman [Groningen (Netherlands)

    2003-10-01

    In the past years in almost all conferences on storage of nuclear waste, ethics has been considered as an important theme. But what is ethics? We will first give a sketch of this branch of philosophy. We will then give a short explanation of the three principal ethical theories. In the discussion about storage of nuclear waste, the ethical theory of utilitarianism is often implicitly invoked. In this system future generations weigh less heavily than the present generation, so that people of the future are not considered as much as those now living. We reject this form of reasoning. The discussion about nuclear waste is also sometimes pursued from ethical points of departure such as equality and justice. But many loose ends remain in these arguments, which gives rise to the question of whether the production and storage of nuclear waste is responsible.

  18. 1993 annual report of hazardous waste activities for the Oak Ridge K-25 site

    Energy Technology Data Exchange (ETDEWEB)

    1994-02-01

    This report is a detailed listing of all of the Hazardous Waste activities occurring at Martin Marietta`s K-25 site. Contained herein are hazardous waste notification forms, waste stream reports, generator fee forms and various TSDR reports.

  19. 1993 annual report of hazardous waste activities for the Oak Ridge K-25 site

    Energy Technology Data Exchange (ETDEWEB)

    1994-02-01

    This report is a detailed listing of all of the Hazardous Waste activities occurring at Martin Marietta`s K-25 site. Contained herein are hazardous waste notification forms, waste stream reports, generator fee forms and various TSDR reports.

  20. An overview of waste crime, its characteristics, and the vulnerability of the EU waste sector.

    Science.gov (United States)

    Baird, J; Curry, R; Cruz, P

    2014-02-01

    While waste is increasingly viewed as a resource to be globally traded, increased regulatory control on waste across Europe has created the conditions where waste crime now operates alongside a legitimate waste sector. Waste crime,is an environmental crime and a form of white-collar crime, which exploits the physical characteristics of waste, the complexity of the collection and downstream infrastructure, and the market opportunities for profit. This paper highlights some of the factors which make the waste sector vulnerable to waste crime. These factors include new legislation and its weak regulatory enforcement, the economics of waste treatment, where legal and safe treatment of waste can be more expensive than illegal operations, the complexity of the waste sector and the different actors who can have some involvement, directly or indirectly, in the movement of illegal wastes, and finally that waste can be hidden or disguised and creates an opportunity for illegal businesses to operate alongside legitimate waste operators. The study also considers waste crime from the perspective of particular waste streams that are often associated with illegal shipment or through illegal treatment and disposal. For each, the nature of the crime which occurs is shown to differ, but for each, vulnerabilities to waste crime are evident. The paper also describes some approaches which can be adopted by regulators and those involved in developing new legislation for identifying where opportunities for waste crime occurs and how to prevent it.

  1. Comparison among the rice bark in the raw and active forms in the removal of {sup 241}Am and {sup 137}Cs from liquid radioactive wastes; Comparacao entre a casca de arroz nas formas brutas e ativada na remocao de {sup 241}Am e {sup 137}Cs de rejeitos radioativos liquidos

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Rafael V.P.; Lima, Josenilson B. de; Bellini, Maria Helena; Sakata, Solange Kazumi; Marumo, Julio Takehiro, E-mail: rpadua@ipen.b, E-mail: sksakata@ipen.b, E-mail: jblima@ipen.b, E-mail: mbmarumo@ipen.b, E-mail: jtmarumo@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-10-26

    New techniques involving treatment of radioactive wastes which associate simplicity and low cost have been directed the attention for the bio sorption, which is a process were solid vegetable or micro-organism for the retention, removing, or recovering of heavy metals from a liquid environment. This study evaluated the capacity of a bio sorbent to remove Am-241 and Cs-137 from liquid radioactive waste. The chosen material was the rice bark employed in the raw or activated forms. The obtained results suggest that the bio sorption, with the activated rice bark, can be a viable technique for the treatment of liquid radioactive wastes containing Am-241 and Cs-137 present in liquid radioactive wastes

  2. Waste management system alternatives for treatment of wastes from spent fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    McKee, R.W.; Swanson, J.L.; Daling, P.M.; Clark, L.L.; Craig, R.A.; Nesbitt, J.F.; McCarthy, D.; Franklin, A.L.; Hazelton, R.F.; Lundgren, R.A.

    1986-09-01

    This study was performed to help identify a preferred TRU waste treatment alternative for reprocessing wastes with respect to waste form performance in a geologic repository, near-term waste management system risks, and minimum waste management system costs. The results were intended for use in developing TRU waste acceptance requirements that may be needed to meet regulatory requirements for disposal of TRU wastes in a geologic repository. The waste management system components included in this analysis are waste treatment and packaging, transportation, and disposal. The major features of the TRU waste treatment alternatives examined here include: (1) packaging (as-produced) without treatment (PWOT); (2) compaction of hulls and other compactable wastes; (3) incineration of combustibles with cementation of the ash plus compaction of hulls and filters; (4) melting of hulls and failed equipment plus incineration of combustibles with vitrification of the ash along with the HLW; (5a) decontamination of hulls and failed equipment to produce LLW plus incineration and incorporation of ash and other inert wastes into HLW glass; and (5b) variation of this fifth treatment alternative in which the incineration ash is incorporated into a separate TRU waste glass. The six alternative processing system concepts provide progressively increasing levels of TRU waste consolidation and TRU waste form integrity. Vitrification of HLW and intermediate-level liquid wastes (ILLW) was assumed in all cases.

  3. Industrial Waste

    DEFF Research Database (Denmark)

    Christensen, Thomas Højlund

    2011-01-01

    Industrial waste is waste from industrial production and manufacturing. Industry covers many industrial sectors and within each sector large variations are found in terms of which raw materials are used, which production technology is used and which products are produced. Available data on unit...... generation rates and material composition as well as determining factors are discussed in this chapter. Characterizing industrial waste is faced with the problem that often only a part of the waste is handled in the municipal waste system, where information is easily accessible. In addition part...... of the industrial waste may in periods, depending on market opportunities and prices, be traded as secondary rawmaterials. Production-specificwaste from primary production, for example steel slag, is not included in the current presentation. In some countries industries must be approved or licensed and as part...

  4. Space disposal of nuclear wastes

    Science.gov (United States)

    Priest, C. C.; Nixon, R. F.; Rice, E. E.

    1980-01-01

    The DOE has been studying several options for nuclear waste disposal, among them space disposal, which NASA has been assessing. Attention is given to space disposal destinations noting that a circular heliocentric orbit about halfway between Earth and Venus is the reference option in space disposal studies. Discussion also covers the waste form, showing that parameters to be considered include high waste loading, high thermal conductivity, thermochemical stability, resistance to leaching, fabrication, resistance to oxidation and to thermal shock. Finally, the Space Shuttle nuclear waste disposal mission profile is presented.

  5. Waste indicators

    Energy Technology Data Exchange (ETDEWEB)

    Dall, O.; Lassen, C.; Hansen, E. [Cowi A/S, Lyngby (Denmark)

    2003-07-01

    The Waste Indicator Project focuses on methods to evaluate the efficiency of waste management. The project proposes the use of three indicators for resource consumption, primary energy and landfill requirements, based on the life-cycle principles applied in the EDIP Project. Trial runs are made With the indicators on paper, glass packaging and aluminium, and two models are identified for mapping the Danish waste management, of which the least extensive focuses on real and potential savings. (au)

  6. Melt processed single phase hollandite waste forms for nuclear waste immobilization: Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, Jake; Marra, James [Savannah River National Laboratory, Aiken, SC 29808 (United States); Conradson, Steven D.; Tang, Ming [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Brinkman, Kyle, E-mail: kyle.brinkman@srnl.doe.gov [Savannah River National Laboratory, Aiken, SC 29808 (United States)

    2014-01-25

    Highlights: • This work shows Cr additions improve the performance of Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al hollandite • Fe–Hol sample exhibited the least Cs retention whereas the Cr–Hol exhibited the greatest Cs retention. • CAF–Hol and Fe–Hol samples exhibited secondary phases enriched in Cs. • X-ray absorption measurements confirmed the stability of Cr{sup +3} as compared to Fe{sup +3}. -- Abstract: Cs is one of the more problematic fission product radionuclides to immobilize due to its high volatility at elevated temperatures, ability to form water soluble compounds, and its mobility in many host materials. The hollandite structure is a promising crystalline host for Cs immobilization and has been traditionally fabricated by solid state sintering methods. This study presents the structure and performance of Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al hollandite fabricated by melt processing. Melt processing is considered advantageous given that melters are currently in use for High Level Waste (HLW) vitrification in several countries. This work details the impact of Cr additions that were demonstrated to (i) promote the formation of a Cs containing hollandite phase and (ii) maintain the stability of the hollandite phase in reducing conditions anticipated for multiphase waste form processing.

  7. Nevada Test Site Waste Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office

    2005-10-01

    This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive (LLW) and mixed waste (MW) for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex (RWMC) for storage or disposal.

  8. Nevada Test Site Waste Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office

    2005-10-01

    This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive (LLW) and mixed waste (MW) for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex (RWMC) for storage or disposal.

  9. 基于结构性双标图的城市废水污染物排放分析%The form biplot analysis of main pollutant emission in waste water in part urban cities

    Institute of Scientific and Technical Information of China (English)

    马艳芳; 张晓琴

    2016-01-01

    Biplot analysis is an useful graphical representation of multivariate date ,which is based on the principal component .Traditional biplot can be divided into three kinds according to the distri-bution of the singular values ,respectively ,covariance biplot ,form biplot ,sym-metric scaling biplot .This paper aim to analyze main pollutant emission in waste water in part urban cities of China by form biplot and get the clustering results .Compared to complete linkage method ,we can conclude that the form biplot not only get the coresponding clusters , but also can get significant variable in these categories .It provide the strong help for wastewater treatment projects .%双标图是一种基于主成分分析原理,将多维数据可视化在一张图中的有效方法。传统的双标图根据奇异值的分配分为协变性双标图、结构性双标图、对称双标图三种。文章主要针对结构性双标图适合分析样本点之间相似性的特点,对全国主要城市的废水污染物排放情况做了双标图分析,并且和最长距离聚类法相比较,结果表明结构性双标图不仅可以得到相应的类别,而且可以得到相应类别中的显著性变量,为废水处理工程提供了有力帮助。

  10. Waste Socio-technological Transitions

    DEFF Research Database (Denmark)

    Zapata Campos, Maria José; Zapata, Patrik; Eriksson-Zetterquist, Ulla

    2014-01-01

    able to replace other production or construction materials. It also means that the consumption of the amount of virgin material decreases and saves energy. Despite the growing material recycling rates, the amount of waste per person, and packaging waste among others, continues to increase. High......The transformation of packaging waste from a problem into a resource has had significant consequences for a more sustainable use of natural resources and even the reduction of potential C02 emissions and its contribution to the climate change. Material recycling leads to separated material being...... recycling rates can be pointless if the amount of waste does not decrease. This is an example of how well established waste recovery regimes can stand in the way of more sustainable forms to handle waste (Corvellec et al, 2013) and, ultimately, hinder the development towards the EU objective (2008...

  11. Food waste or wasted food

    OpenAIRE

    van Graas, Maaike Helene

    2014-01-01

    In the industrialized world large amounts of food are daily disposed of. A significant share of this waste could be avoided if different choices were made by individual households. Each day, every household makes decisions to maximize their happiness while balancing restricted amounts of time and money. Thinking of the food waste issue in terms of the consumer choice problem where households can control the amount of wasted food, we can model how households can make the best decisions. I...

  12. Engineered waste-package-system design specification

    Energy Technology Data Exchange (ETDEWEB)

    1983-05-01

    This report documents the waste package performance requirements and geologic and waste form data bases used in developing the conceptual designs for waste packages for salt, tuff, and basalt geologies. The data base reflects the latest geotechnical information on the geologic media of interest. The parameters or characteristics specified primarily cover spent fuel, defense high-level waste, and commercial high-level waste forms. The specification documents the direction taken during the conceptual design activity. A separate design specification will be developed prior to the start of the preliminary design activity.

  13. Radioactive waste material melter apparatus

    Science.gov (United States)

    Newman, D.F.; Ross, W.A.

    1990-04-24

    An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs.

  14. 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.

  15. Charge-coupled substituted garnets (Y3-xCa0.5xM0.5x)Fe5O12 (M = Ce, Th): structure and stability as crystalline nuclear waste forms.

    Science.gov (United States)

    Guo, Xiaofeng; Kukkadapu, Ravi K; Lanzirotti, Antonio; Newville, Matthew; Engelhard, Mark H; Sutton, Stephen R; Navrotsky, Alexandra

    2015-04-20

    The garnet structure has been proposed as a potential crystalline nuclear waste form for accommodation of actinide elements, especially uranium (U). In this study, yttrium iron garnet (YIG) as a model garnet host was studied for the incorporation of U analogs, cerium (Ce) and thorium (Th), incorporated by a charge-coupled substitution with calcium (Ca) for yttrium (Y) in YIG, namely, 2Y(3+) = Ca(2+) + M(4+), where M(4+) = Ce(4+) or Th(4+). Single-phase garnets Y3-xCa0.5xM0.5xFe5O12 (x = 0.1-0.7) were synthesized by the citrate-nitrate combustion method. Ce was confirmed to be tetravalent by X-ray absorption spectroscopy and X-ray photoelectron spectroscopy. X-ray diffraction and (57)Fe-Mössbauer spectroscopy indicated that M(4+) and Ca(2+) cations are restricted to the c site, and the local environments of both the tetrahedral and the octahedral Fe(3+) are systematically affected by the extent of substitution. The charge-coupled substitution has advantages in incorporating Ce/Th and in stabilizing the substituted phases compared to a single substitution strategy. Enthalpies of formation of garnets were obtained by high temperature oxide melt solution calorimetry, and the enthalpies of substitution of Ce and Th were determined. The thermodynamic analysis demonstrates that the substituted garnets are entropically rather than energetically stabilized. This suggests that such garnets may form and persist in repositories at high temperature but might decompose near room temperature.

  16. Charge-coupled Substituted Garnets (Y3-xCa0.5xM0.5x)Fe5O12 (M = Ce, Th): Structure and Stability as Crystalline Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Xiaofeng; Kukkadapu, Ravi K.; Lanzirotti, Anthony; Newville, Mathew; Engelhard, Mark H.; Sutton , Steven R.; Navrotsky, Alexandra

    2015-04-20

    The garnet structure has been proposed as a potential crystalline nuclear waste form for accommodation of actinide elements, especially uranium (U). In this study, yttrium iron garnet (YIG) as a model garnet host was studied for the incorporation of U analogs, cerium (Ce), and thorium (Th), incorporated by a charge-coupled substitution with calci-um (Ca) for yttrium (Y) in YIG, namely 2Y3+ = Ca2+ + M4+, where M4+ = Ce4+ or Th4+. Single phase garnets Y3-xCa0.5xM0.5xFe5O12, synthesized by the citrate-nitrate combustion method, were obtained up to x = 0.7. Ce was confirmed to be tetravalent by X-ray absorption spectroscopy and X-ray photoelectron spectroscopy. X-ray diffraction and 57Fe-Mössbauer spectroscopy indicated that the samples are single phase, M4+ and Ca2+ cations are restricted to the c-site, the nature of M4+ has only a minor effect on the structure, and the local environments of both the tetrahedral and octahedral Fe3+ are systematically affected by the extent of substitution, especially on the tetrahedral sublattice. The charge coupled substitution has advantages in incorporating Ce/Th and in stabilizing the substituted phases, compared to a single substitution strategy. Enthalpies of formation of garnets were obtained by high temperature oxide melt solution calorimetry, and the enthalpies of substitution of Ce and Th were determined. The thermodynamic analysis demonstrates that the substituted garnets are entropically rather than energetically stabilized. This suggests that such garnets may form and persist in repositories at high temperature but might decompose near room temperature. These structural and thermodynamic findings shed light on possible incorporation of U in this garnet system.

  17. Hazardous-waste analysis plan for LLNL operations

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, R.S.

    1982-02-12

    The Lawrence Livermore National Laboratory is involved in many facets of research ranging from nuclear weapons research to advanced Biomedical studies. Approximately 80% of all programs at LLNL generate hazardous waste in one form or another. Aside from producing waste from industrial type operations (oils, solvents, bottom sludges, etc.) many unique and toxic wastes are generated such as phosgene, dioxin (TCDD), radioactive wastes and high explosives. One key to any successful waste management program must address the following: proper identification of the waste, safe handling procedures and proper storage containers and areas. This section of the Waste Management Plan will address methodologies used for the Analysis of Hazardous Waste. In addition to the wastes defined in 40 CFR 261, LLNL and Site 300 also generate radioactive waste not specifically covered by RCRA. However, for completeness, the Waste Analysis Plan will address all hazardous waste.

  18. Collaborative form(s)

    DEFF Research Database (Denmark)

    Gunn, Wendy

    Gunn asks us to consider beauty as collaborative forms of action generated by moving between design by means of anthropology and anthropology by means of design. Specifically, she gives focus to play-like reflexions on practices of designing energy products, systems and infrastructure. Design...

  19. Health and Environmental Hazards of Electronic Waste in India.

    Science.gov (United States)

    Borthakur, Anwesha

    2016-04-01

    Technological waste in the form of electronic waste (e-waste) is a threat to all countries. E-waste impacts health and the environment by entering the food chain in the form of chemical toxicants and exposing the population to deleterious chemicals, mainly in the form of polycyclic aromatic hydrocarbons and persistent organic pollutants. This special report tries to trace the environmental and health implications of e-waste in India. The author concludes that detrimental health and environmental consequences are associated with e-waste and the challenge lies in producing affordable electronics with minimum chemical toxicants.

  20. Nevada National Security Site Waste Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Environmental Management

    2010-09-03

    This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

  1. Nevada National Security Site Waste Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Environmental Management

    2011-01-01

    This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

  2. 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.

  3. Waste acceptance criteria for the Waste Isolation Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-04-01

    The Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC), DOE/WIPP-069, was initially developed by a U.S. Department of Energy (DOE) Steering Committee to provide performance requirements to ensure public health and safety as well as the safe handling of transuranic (TRU) waste at the WIPP. This revision updates the criteria and requirements of previous revisions and deletes those which were applicable only to the test phase. The criteria and requirements in this document must be met by participating DOE TRU Waste Generator/Storage Sites (Sites) prior to shipping contact-handled (CH) and remote-handled (RH) TRU waste forms to the WIPP. The WIPP Project will comply with applicable federal and state regulations and requirements, including those in Titles 10, 40, and 49 of the Code of Federal Regulations (CFR). The WAC, DOE/WIPP-069, serves as the primary directive for assuring the safe handling, transportation, and disposal of TRU wastes in the WIPP and for the certification of these wastes. The WAC identifies strict requirements that must be met by participating Sites before these TRU wastes may be shipped for disposal in the WIPP facility. These criteria and requirements will be reviewed and revised as appropriate, based on new technical or regulatory requirements. The WAC is a controlled document. Revised/changed pages will be supplied to all holders of controlled copies.

  4. Nuclear waste management. Quarterly progress report, April-June 1981

    Energy Technology Data Exchange (ETDEWEB)

    Chikalla, T.D.; Powell, J.A.

    1981-09-01

    Reports and summaries are presented for the following: high-level waste process development; alternative waste forms; TMI zeolite vitrification demonstration program; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton implantation; thermal outgassing; iodine-129 fixation; NWVP off-gas analysis; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; verification instrument development; mobility of organic complexes of radionuclides in soils; handbook of methods to decrease the generation of low-level waste; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology program; high-level waste form preparation; development of backfill materials; development of structural engineered barriers; disposal charge analysis; and analysis of spent fuel policy implementation.

  5. Communication from the Radioactive Waste Service

    CERN Multimedia

    2011-01-01

    The Radioactive Waste service of the Radiation protection Group informs you that as of 15 April 2011 radioactive waste can be delivered to the waste treatment centre (Bldg. 573) only during the following hours: Mon- Thu: 08:00 – 11:30 / 13:30 – 16:00 Fri : 08:00 – 11:30 An electronic form must be filled in before the arrival of the waste at the treatment centre: https://edh.cern.ch/Document/General/RadioactiveWaste for further information, please call 73171.

  6. Study on α-Form Calcuim Sulfate Hemhydra from Flue Gas Desulfuriza Scrubbing Sludge Waste%从烟气脱硫残渣中制取α-半水石膏的研究

    Institute of Scientific and Technical Information of China (English)

    胥桂萍; 童仕唐

    2001-01-01

    对溶液法从烟气脱硫残渣中制取α半水石膏进行了进一步的研究。FGD残渣在碱土金属盐类的盐溶液中,在丁二酸、丙三醇等结晶改良剂的作用下,与大气压下加热脱水生成α半水石膏。研究中发现pH值和盐溶液浓度是影响FGD石膏脱水速度最敏感的因素,固液比虽然对脱水速度影响不大,但固液比增大对晶体生长和习性改良不利。pH值愈接近中性范围,愈能增强α半水石膏结晶在液相中的稳定性,有利于结晶习性改良。%Some study are done for the new process for conversion of flue gas desulfurization scrubbing sludge waste. FGD gypsum further into α-form calcium sulfate hemihydrate by heating in an aqueous solution consisting of some alkail earth metal salts and sulfuric acid or chlorhydric acid at the atmospheric pressure has been studied in this subject. It is found that the concentration of sulfuric acid in liquid and the reaction temperature are the most sensitive factors for this process. Increasing the ratio of solid to liquid is disadvantageous for growth of crystals even though it does not effect obviously for the rate of dehydration of FGD gypsum. On the other hand,the more closed to the neutral range of pH value the liquid is adjusted,the better stability of the α-form calcium sulfate hemihydrate appear,the more favorable for modification of crystal habit the process is made.

  7. Study of the dechroming of tanned leather wastes

    OpenAIRE

    Botić Tatjana; Ilišković Nadežda V.; Drljača Dijana

    2004-01-01

    According to European legislation, it is not possible to dump any chromium-containing waste in Europe. The minimization of wastes is a key element in that strategy. It involves the application of clean technologies: low and non-waste technologies. The tanning industry generates substantial quantities of chromium-containing solid waste in the form of shavings and trimmings. The recycling and reuse of those wastes must be the primary target in optimizing processes of the leather industry. The p...

  8. Tritium waste package

    Science.gov (United States)

    Rossmassler, Rich; Ciebiera, Lloyd; Tulipano, Francis J.; Vinson, Sylvester; Walters, R. Thomas

    1995-01-01

    A containment and waste package system for processing and shipping tritium xide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen add oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB.

  9. Landfills - Municipal Waste Operations

    Data.gov (United States)

    NSGIC GIS Inventory (aka Ramona) — A Municipal Waste Operation is a DEP primary facility type related to the Waste Management Municipal Waste Program. The sub-facility types related to Municipal Waste...

  10. Closed Fuel Cycle Waste Treatment Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, J. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Collins, E. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Crum, J. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, S. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Garn, T. G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gombert, D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maio, V. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Matyas, J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Nenoff, T. M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Riley, B. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sevigny, G. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thallapally, P. K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, J. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-02-01

    This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form

  11. Resource Conservation and Recovery Act, Part B Permit Application [for the Waste Isolation Pilot Plant (WIPP)]. Volume 2, Chapter C, Appendix C1--Chapter C, Appendix C3 (beginning), Revision 3

    Energy Technology Data Exchange (ETDEWEB)

    1993-03-01

    This volume contains appendices for the following: Rocky Flats Plant and Idaho National Engineering Laboratory waste process information; TRUPACT-II content codes (TRUCON); TRUPACT-II chemical list; chemical compatibility analysis for Rocky Flats Plant waste forms; chemical compatibility analysis for waste forms across all sites; TRU mixed waste characterization database; hazardous constituents of Rocky Flats Transuranic waste; summary of waste components in TRU waste sampling program at INEL; TRU waste sampling program; and waste analysis data.

  12. Human waste

    NARCIS (Netherlands)

    Amin, Md Nurul; Kroeze, Carolien; Strokal, Maryna

    2017-01-01

    Many people practice open defecation in south Asia. As a result, lot of human waste containing nutrients such as nitrogen (N) and phosphorus (P) enter rivers. Rivers transport these nutrients to coastal waters, resulting in marine pollution. This source of nutrient pollution is, however, ignored in

  13. Waste disposal

    CERN Multimedia

    2006-01-01

    We should like to remind you that you can have all commonplace, conventional waste (combustible, inert, wood, etc.) disposed of by the TS-FM Group. Requests for the removal of such waste should be made by contacting FM Support on tel. 77777 or by e-mail (Fm.Support@cern.ch). For requests to be acted upon, the following information must be communicated to FM Support: budget code to be debited for the provision and removal of the skip / container. type of skip required (1m3, 4 m3, 7 m3, 15 m3, 20 m3, 30 m3). nature of the waste to be disposed of (bulky objects, cardboard boxes, etc.). building concerned. details of requestor (name, phone number, department, group, etc.). We should also like to inform you that the TS-FM Group can arrange for waste to be removed from work-sites for firms under contract to CERN, provided that the prior authorisation of the CERN Staff Member in charge of the contract is obtained and the relevant disposal/handling charges are paid. You are reminded that the selective sorting o...

  14. Waste disposal

    CERN Multimedia

    2006-01-01

    We should like to remind you that you can have all commonplace, conventional waste (combustible, inert, wood, etc.) disposed of by the TS-FM Group. Requests for the removal of such waste should be made by contacting FM Support on tel. 77777 or by e-mail (Fm.Support@cern.ch). For requests to be acted upon, the following information must be communicated to FM Support: budget code to be debited for the provision and removal of the skip / container; type of skip required (1m3, 4 m3, 7 m3, 15 m3, 20 m3, 30 m3); nature of the waste to be disposed of (bulky objects, cardboard boxes, etc.); building concerned; details of requestor (name, phone number, department, group, etc.). We should also like to inform you that the TS-FM Group can arrange for waste to be removed from work-sites for firms under contract to CERN, provided that the prior authorisation of the CERN Staff Member in charge of the contract is obtained and the relevant disposal/handling charges are paid. You are reminded that the selective sorting...

  15. Producing a solid fuel from agricultural wastes

    Energy Technology Data Exchange (ETDEWEB)

    Khigasikuni, T.; Fudziki, A.; Koisi, K.

    1982-08-21

    Agricultural wastes, in particular, the peels and seeds of mandarine oranges, used for canning, were ground, dried, mixed with a binder, molded in cylindrical or pyramidal forms of a piece of the desired size, heated and impregnated with a water repellant substance, for instance, paraffin. A material is produced with a d = 1. PVA was used as the binder. The wastes were partially gasified and the obtained gas was used as a heat source in drying the wastes.

  16. Tank Waste Disposal Program redefinition

    Energy Technology Data Exchange (ETDEWEB)

    Grygiel, M.L.; Augustine, C.A.; Cahill, M.A.; Garfield, J.S.; Johnson, M.E.; Kupfer, M.J.; Meyer, G.A.; Roecker, J.H. [Westinghouse Hanford Co., Richland, WA (United States); Holton, L.K.; Hunter, V.L.; Triplett, M.B. [Pacific Northwest Lab., Richland, WA (United States)

    1991-10-01

    The record of decision (ROD) (DOE 1988) on the Final Environmental Impact Statement, Hanford Defense High-Level, Transuranic and Tank Wastes, Hanford Site, Richland Washington identifies the method for disposal of double-shell tank waste and cesium and strontium capsules at the Hanford Site. The ROD also identifies the need for additional evaluations before a final decision is made on the disposal of single-shell tank waste. This document presents the results of systematic evaluation of the present technical circumstances, alternatives, and regulatory requirements in light of the values of the leaders and constitutents of the program. It recommends a three-phased approach for disposing of tank wastes. This approach allows mature technologies to be applied to the treatment of well-understood waste forms in the near term, while providing time for the development and deployment of successively more advanced pretreatment technologies. The advanced technologies will accelerate disposal by reducing the volume of waste to be vitrified. This document also recommends integration of the double-and single-shell tank waste disposal programs, provides a target schedule for implementation of the selected approach, and describes the essential elements of a program to be baselined in 1992.

  17. Radioactive waste management in Austria

    Directory of Open Access Journals (Sweden)

    Neubauer Josef

    2004-01-01

    Full Text Available At the Austrian Research Centers Seibersdorf, there are several facilities in stalled for treatment of waste of low and intermediate radioactivity level (radwaste. A separate company within Centers, Nuclear Engineering Seibersdorf, has been formed recently, acting as a centralized facility for treatment, conditioning and storing of such waste within the country. The relevant treatment technology is applied depending on the waste category. In total about 6900 m3 of solid waste of low and intermediate radioactivity level originating from Austria was treated in the period between 1976 and 2002. Presently, there exists no final repository for radwaste in Austria. A study is under way to identify the structure for a long term storage facility.

  18. Wasting and stunting - similarities and differences

    DEFF Research Database (Denmark)

    Briend, André; Khara, Tanya; Dolan, Carmel

    2015-01-01

    Wasting and stunting are often presented as two separate forms of malnutrition requiring different interventions for prevention and/or treatment. These two forms of malnutrition, however, are closely related and often occur together in the same populations and often in the same children. Wasting...... and stunting are both associated with increased mortality, especially when both are present in the same child. A better understanding of the pathophysiology of these two different forms of malnutrition is needed to design efficient programs. A greatly reduced muscle mass is characteristic of severe wasting...... that to decrease malnutrition-related mortality, interventions should aim at preventing both wasting and stunting, which often share common causes. Also, this suggests that treatment interventions should focus on children who are both wasted and stunted and therefore have the greatest deficits in muscle mass...

  19. Phase Stability Determinations of DWPF Waste Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Marra, S.L.

    1999-10-22

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. To fulfill this requirement, glass samples were heat treated at various times and temperatures. These results will provide guidance to the repository program about conditions to be avoided during shipping, handling and storage of DWPF canistered waste forms.

  20. Radiological, physical, and chemical characterization of transuranic wastes stored at the Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Apel, M.L.; Becker, G.K.; Ragan, Z.K.; Frasure, J.; Raivo, B.D.; Gale, L.G.; Pace, D.P.

    1994-03-01

    This document provides radiological, physical and chemical characterization data for transuranic radioactive wastes and transuranic radioactive and hazardous (i.e., mixed) wastes stored at the Idaho National Engineering Laboratory and considered for treatment under the Private Sector Participation Initiative Program (PSPI). Waste characterization data are provided in the form of INEL Waste Profile Sheets. These documents provide, for each content code, information on waste identification, waste description, waste storage configuration, physical/chemical waste composition, radionuclide and associated alpha activity waste characterization data, and hazardous constituents present in the waste. Information is provided for 139 waste streams which represent an estimated total volume of 39,380{sup 3} corresponding to a total mass of approximately 19,000,000 kg. In addition, considerable information concerning alpha, beta, gamma, and neutron source term data specific to Rocky Flats Plant generated waste forms stored at the INEL are provided to assist in facility design specification.

  1. Rethinking the waste hierarchy

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, C.; Vigsoe, D. (eds.)

    2005-03-01

    There is an increasing need to couple environmental and economic considerations within waste management. Consumers and companies alike generate ever more waste. The waste-policy challenges of the future lie in decoupling growth in waste generation from growth in consumption, and in setting priorities for the waste management. This report discusses the criteria for deciding priorities for waste management methods, and questions the current principles of EU waste policies. The basis for the discussion is the so-called waste hierarchy which has dominated the waste policy in the EU since the mid-1970s. The waste hierarchy ranks possible methods of waste management. According to the waste hierarchy, the very best solution is to reduce the amount of waste. After that, reuse is preferred to recycling which, in turn, is preferred to incineration. Disposal at a landfill is the least favourable solution. (BA)

  2. Other Special Waste

    DEFF Research Database (Denmark)

    Brogaard, Line Kai-Sørensen; Christensen, Thomas Højlund

    2011-01-01

    separately from MSW. Some of these other special wastes are briefly described in this chapter with respect to their definition, quantity and composition, and management options. The special wastes mentioned here are batteries, tires, polyvinylchloride (PVC) and food waste.......In addition to the main types of special waste related to municipal solid waste (MSW) mentioned in the previous chapters (health care risk waste, WEEE, impregnated wood, hazardous waste) a range of other fractions of waste have in some countries been defined as special waste that must be handled...

  3. Other Special Waste

    DEFF Research Database (Denmark)

    Brogaard, Line Kai-Sørensen; Christensen, Thomas Højlund

    2011-01-01

    In addition to the main types of special waste related to municipal solid waste (MSW) mentioned in the previous chapters (health care risk waste, WEEE, impregnated wood, hazardous waste) a range of other fractions of waste have in some countries been defined as special waste that must be handled...... separately from MSW. Some of these other special wastes are briefly described in this chapter with respect to their definition, quantity and composition, and management options. The special wastes mentioned here are batteries, tires, polyvinylchloride (PVC) and food waste....

  4. Development of polyphase ceramics for the immobilization of high-level Defense nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, P.E.D.; Harker, A.B.; Clarke, D.R.; Flintoff, J.J.; Shaw, T.M.

    1983-02-25

    The report contains two major sections: Section I - An Improved Polyphase Ceramic for High-Level Defense Nucleation Waste reports the work conducted on titanium-silica based ceramics for immobilizing Savannah River Plant waste. Section II - Formulation and Processing of Alumina Based Ceramic Nuclear Waste Forms describes the work conducted on developing a generic alumina and alumina-silica based ceramic waste form capable of immobilizing any nuclear waste with a high aluminum content. Such wastes include the Savannah River Plant wastes, Hanford neutralized purex wastes, and Hanford N-Reactor acid wastes. The design approach and process technology in the two reports demonstrate how the generic high waste loaded ceramic form can be applied to a broad range of nuclear waste compositions. The individual sections are abstracted and indexed separately.

  5. Remaining Sites Verification Package for the 100-B-21:2 Subsite (100-B/C Discovery Pipeline DS-100BC-002), Waste Site Reclassification Form 2008-003

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-06-16

    The 100-B-21:2 waste site consists of the immediate area of the DS-100BC-02 pipeline. In accordance with this evaluation, the confirmatory and verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  6. Hanford Site annual dangerous waste report: Volume 1, Part 1, Generator dangerous waste report, dangerous waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    This report contains information on hazardous wastes at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, waste number, weight, and waste designation.