WorldWideScience

Sample records for stable waste form

  1. Manifolds admitting stable forms

    Czech Academy of Sciences Publication Activity Database

    Le, Hong-Van; Panák, Martin; Vanžura, Jiří

    2008-01-01

    Roč. 49, č. 1 (2008), s. 101-11 ISSN 0010-2628 R&D Projects: GA ČR(CZ) GP201/05/P088 Institutional research plan: CEZ:AV0Z10190503 Keywords : stable forms * automorphism groups Subject RIV: BA - General Mathematics

  2. Waste forms for plutonium disposition

    International Nuclear Information System (INIS)

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

    1997-01-01

    The field of plutonium disposition is varied and of much importance, since the Department of Energy has decided on the hybrid option for disposing of the weapons materials. This consists of either placing the Pu into mixed oxide fuel for reactors or placing the material into a stable waste form such as glass. The waste form used for Pu disposition should exhibit certain qualities: (1) provide for a suitable deterrent to guard against proliferation; (2) be of minimal volume, i.e., maximize the loading; and (3) be reasonably durable under repository-like conditions. This paper will discuss several Pu waste forms that display promising characteristics

  3. Waste-form development

    International Nuclear Information System (INIS)

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

    1982-01-01

    Contemporary solidification agents are being investigated relative to their applications to major fuel cycle and non-fuel cycle low-level waste (LLW) streams. Work is being conducted to determine the range of conditions under which these solidification agents can be applied to specific LLW streams. These studies are directed primarily towards defining operating parameters for both improved solidification of problem wastes and solidification of new LLW streams generated from advanced volume reduction technologies. Work is being conducted to measure relevant waste form properties. These data will be compiled and evaluated to demonstrate compliance with waste form performance and shallow land burial acceptance criteria and transportation requirements

  4. Comparative waste forms study

    International Nuclear Information System (INIS)

    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

  5. Waste form development

    International Nuclear Information System (INIS)

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

    1982-01-01

    In this program, contemporary solidification agents are being investigated relative to their applications to major fuel cycle and non-fuel cycle low-level waste (LLW) streams. Work is being conducted to determine the range of conditions under which these solidification agents can be applied to specific LLW streams. These studies are directed primarily towards defining operating parameters for both improved solidification of problem wastes and solidification of new LLW streams generated from advanced volume reduction technologies. Work is being conducted to measure relevant waste form properties. These data will be compiled and evaluated to demonstrate compliance with waste form performance and shallow land burial acceptance criteria and transportation requirements (both as they exist and as they are modified with time). 6 tables

  6. Transuranic waste management program waste form development

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  7. 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.

  8. DWPF waste form compliance plan (Draft Revision)

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  9. Advanced waste forms from spent nuclear fuel

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  10. Mixed Waste Focus Area - Waste form initiative

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  11. Survey of concrete waste forms

    International Nuclear Information System (INIS)

    Moore, J.G.

    1981-01-01

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

  12. Stainless steel-zirconium alloy waste forms

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  13. Waste form development/test

    International Nuclear Information System (INIS)

    Kalb, P.D.; Colombo, P.

    1983-01-01

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

  14. Coated particle waste form development

    Energy Technology Data Exchange (ETDEWEB)

    Oma, K.H.; Buckwalter, C.Q.; Chick, L.A.

    1981-12-01

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

  15. Coated particle waste form development

    International Nuclear Information System (INIS)

    Oma, K.H.; Buckwalter, C.Q.; Chick, L.A.

    1981-12-01

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

  16. 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

  17. TSA waste stream and final waste form composition

    International Nuclear Information System (INIS)

    Grandy, J.D.; Eddy, T.L.; Anderson, G.L.

    1993-01-01

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

  18. Nuclear waste forms for actinides

    Science.gov (United States)

    Ewing, Rodney C.

    1999-01-01

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

  19. 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.

  20. Processes for production of alternative waste forms

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  1. 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.

  2. Waste Disposal: R and D on Waste Forms and Packages

    International Nuclear Information System (INIS)

    Van Iseghem, P.

    2000-01-01

    The main objectives of SCK-CEN's programme on waste forms and waste packages are: (1) to determine or to verify various physical and chemical characteristics of radioactive waste forms relevant to the Belgian waste management programme; (2) to improve and to develop tools, methods and approaches for characterising radioactive waste; (3) to assess experimentally or to demonstrate in situ the long-term performance of radioactive waste forms with regard to geological disposal in clay; (4) to assess the performance of candidate overpack materials through in situ and laboratory experiments

  3. Miscellaneous Waste-Form FEPs

    International Nuclear Information System (INIS)

    Schenker, A.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Crandall, J.L.

    1979-01-01

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

  5. Alternative solidified forms for nuclear wastes

    International Nuclear Information System (INIS)

    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

  6. 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

  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); 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.

  8. A generalized definition for waste form durability

    International Nuclear Information System (INIS)

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

    2002-01-01

    When evaluating waste form performance, the term ''durability'' often appears in casual discourse, but in the technical literature, the focus is often on waste form ''degradation'' in terms of mass lost per unit area per unit time. Waste form degradation plays a key role in developing models of the long-term performance in a repository environment, but other factors also influence waste form performance. These include waste form geometry; density, porosity, and cracking; the presence of cladding; in-package chemistry feedback; etc. The paper proposes a formal definition of waste form ''durability'' which accounts for these effects. Examples from simple systems as well as from complex models used in the Total System Performance Assessment of Yucca Mountain are provided. The application of ''durability'' in the selection of bounding models is also discussed

  9. 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.

  10. Waste Form Evaluation Program. Final report

    International Nuclear Information System (INIS)

    Franz, E.M.; Colombo, P.

    1985-09-01

    This report presents data that can be used to assess the acceptability of polyethylene and modified sulfur cement waste forms to meet the requirements of 10 CFR 61. The waste streams selected for this study include dry evaporator concentrate salts and incinerator ash as representative wastes which result from advanced volume reduction technologies and ion exchange resins which remain problematic for solidification using commercially available matrix materials. Property evaluation tests such as compressive strength, water immersion, thermal cycling, irradiation, biodegradation and leachability were conducted for polyethylene and sulfur cement waste forms over a range of waste-to-binder ratios. Based on the results of the tests, optimal waste loadings of 70 wt % sodium sulfate, 50 wt % boric acid, 40 wt % incinerator ash and 30 wt % ion exchange resins were established for polyethylene, although maximum loadings were considerably higher. For modified sulfur cement, optimal loadings of 40 wt % sodium sulfate, 40 wt % boric acid and 40 wt % incinerator ash are reported. Ion exchange resins are not recommended for incorporation into modified sulfur cement because of poor waste form performance even at very low waste concentrations. The results indicate that all waste forms tested within the range of optimal waste concentrations satisifed the requirements of the NRC Technical Position Paper on Waste Form

  11. Secondary waste form testing: ceramicrete phosphate bonded ceramics

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. 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

  13. Summary: special waste form lysimeters - arid program

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  14. Summary and evaluation of nuclear waste forms

    International Nuclear Information System (INIS)

    Lutze, W.; Ewing, R.C.

    1988-01-01

    Typically, the final comparison of waste forms (e.g., Hench et al. 1984) rested on the predicted risk to man, often calculated as a dose-to-man over long periods of time (Cheung et al. 1982). The codes (e.g., MISER) used in these predictions calculated the transport behavior of radionuclides from the repository and, not surprisingly, found that the differences in the performance between waste forms (e.g., glass versus ceramic) did not result in significant differences in the final ''dose-to-man''. Such conclusions formed the basis for the discontinuation of research programs on alternative or second generation waste forms. This was most unfortunate, because despite the seemingly quantitative results of such calculations, the codes did not take advantage of a detailed knowledge of the various process by which waste form might degrade (e.g., fracture, corrosion or radiation effects). Thus, the conclusion that the calculated ''dose-to-man'' was not sensitive to the orders of magnitude variation in the release of radionuclides from different types of waste forms was not based on any scientific understanding of waste form degradation processes. Significantly, such calculations obscured important differences in the materials properties among waste forms and the different ways in which waste forms interact with various geologic environments. In this chapter, the authors have tried to avoid an approach of this type and have instead compiled data from the contributed chapters into tables

  15. Leaching of nuclear power reactor wastes forms

    International Nuclear Information System (INIS)

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

    1986-01-01

    The leaching tests for power reactor wastes carried out at IPEN/CNEN-SP are described. These waste forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. 3 years leaching results are reported, determining cesium and strontium diffusivity coefficients for boric acid waste form and ion-exchange resins. (Author) [pt

  16. Laboratory procedures for waste form testing

    International Nuclear Information System (INIS)

    Mast, E.S.

    1994-01-01

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

  17. Laboratory procedures for waste form testing

    Energy Technology Data Exchange (ETDEWEB)

    Mast, E.S.

    1994-09-19

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

  18. 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.

  19. 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

  20. Leaching of nuclear power reactor waste forms

    International Nuclear Information System (INIS)

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

    1987-01-01

    The leaching tests for immobilized power reactor wastes carried out at IPEN are described. These wastes forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. Three years leaching results are reported. The cesium diffuvity coefficients determined out of these results are about 1 x 10 -8 cm 2 /s for boric acid waste form and 9 x 10 -9 cm 2 /s for ion-exchange resin waste. Strontium diffusivity coefficients found are about 3 x 10 -11 cm 2 /s and 9 x 10 -11 cm 2 /s respectively. (Author) [pt

  1. 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.

  2. Mixed low-level waste form evaluation

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  3. Development and characterization of cermet forms for radioactive waste

    International Nuclear Information System (INIS)

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

    1979-01-01

    Cermets designed to isolate high-level wastes in a solid form are a composite consisting of various ceramic phase particles uniformly dispersed in and microencapsulated by an iron-nickel base alloy matrix. The metal matrix provides this waste form with many advantageous features including excellent thermal conductivity and mechanical strength. These cermets are formed by first dissolving the waste in molten urea, precipitating and calcining all the constituents, compacting the calcine, and sintering and reduction to form the final product. The exact formulation of cermets through additions to the waste is designed to fix most of the fission products in stable, leach resistant ceramic phases which are subsequently microencapsulated by an alloy matrix. The alloy matrix, which is derived primarily from the waste itself and includes the reducible fission and activation products from the waste, can be compositionally adjusted through additions to optimize its corrosion resistance under conditions existing in various disposal environments. The processes by which cermets are formed include several new and unique materials preparation options that are being developed to permit engineering scale-up and to be compatible with remote operations. Cermets formed by alternate processing methods are being characterized. Initially, cermet samples were prepared using a laboratory scale, batch process developed for the preparation of special ceramics having high compositional uniformity and excellent sinterability. The modification of this batch process to one suitable for scale-up and remote operation is the subject of this paper. Cermet characterization is also discussed

  4. Stability of High-Level Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, Theodore M.; Vienna, John D.

    2005-09-30

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

  5. 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

  6. Igneous Intrusion Impacts on Waste Packages and Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    P. Bernot

    2004-08-16

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

  7. 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.

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

    International Nuclear Information System (INIS)

    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

  9. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    This report describes the scientific basis underlying the approach being followed to design and develop ''advanced'' glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na 2 O 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

  11. 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

  12. Processability analysis of candidate waste forms

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  13. Waste form qualification program review

    International Nuclear Information System (INIS)

    Saverot, P.; Sombret, C.

    1987-01-01

    To meet waste acceptance specifications, it is necessary to demonstrate that the glass produced at the site is suitable for long-term disposal. The experimental studies carried out in France for such a demonstration have been performed according to the following strategy: (1) selection, definition, and characterization of a reference glass; (2) performance of sensitivity tests to quality a sufficiently broad compositional region that covers all credible compositions to be produced; (3) demonstration that the inevitable process disturbances taking place in the glass-making process can be detected and corrected in time to avoid any additional modification of glass formulations from those obtained in the range tested during the sensitivity test; and (4) implementation of a quality control plan guaranteeing the control of the process parameters under normal and degraded operating conditions

  14. 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.

  15. Characterization of radioactive waste forms and packages

    International Nuclear Information System (INIS)

    1997-01-01

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

  16. Lysimeter tests of SRP waste forms

    International Nuclear Information System (INIS)

    Hooker, R.L.; Root, R.W. Jr.

    1981-05-01

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

  17. 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

  18. Determining leach rates of monolithic waste forms

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  19. Construction of solid waste form test facility

    International Nuclear Information System (INIS)

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

    1989-02-01

    The Solid Waste Form Test Facility (SWFTF) is now construction at DAEDUCK in Korea. In SWFTF, the characteristics of solidified waste products as radiological homogeneity, mechanical and thermal property, water resistance and lechability will be tested and evaluated to meet conditions for long-term storage or final disposal of wastes. The construction of solid waste form test facility has been started with finishing its design of a building and equipments in Sep. 1984, and now building construction is completed. Radioactive gas treatment system, extinguishers, cooling and heating system for the facility, electrical equipments, Master/Slave manipulator, power manipulator, lead glass and C.C.T.V. has also been installed. SWFTF will be established in the beginning of 1990's. At this report, radiation shielding door, nondestructive test of the wall, instrumentation system for the utility supply system and cell lighting system are described. (Author)

  20. Preparation techniques for ceramic waste form powder

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  1. 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.

  2. 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.

  3. Low temperatue ceramic radioactive waste form

    International Nuclear Information System (INIS)

    Roy, D.M.; Scheetz, B.E.; Grutzeck, M.W.; Sarkar, A.K.; Atkinson, S.D.

    1979-01-01

    Preliminary research on low temperature ceramic waste forms based upon modified calcium silicate and aluminate cements is described. Compositional and reaction variations, modified to achieve a strongly consolidated waste form which is reistant to significant change under the anticipated ambient, have been investigated. Extending previous studies in this laboratory, the recent work has utilized combinations of calcium aluminate cements and high early strength portland cements, different supercalcine waste formulations, and other additives. Both modified conventional cement composite consolidation methods and low-temperature hot-pressing methods were used to incorporate the supercalcine into a consolidated form. Products after initial consolidation and after drastic accelerated leaching treatment (hydrothermal leaching) were characterized. Uranium and simulated transuranics were found to be highly resistant to leaching; strontium was relatively resistant in many cases; and cesium leaching was found to be compositionally dependent. 8 tables

  4. High-level waste-form-product performance evaluation

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.; Allender, J.S.; Stone, J.A.; Gordon, D.E.; Gould, T.H. Jr.; Westberry, C.F. III.

    1982-01-01

    Seven candidate waste forms were evaluated for immobilization and geologic disposal of high-level radioactive wastes. The waste forms were compared on the basis of leach resistance, mechanical stability, and waste loading. All forms performed well at leaching temperatures of 40, 90, and 150 0 C. Ceramic forms ranked highest, followed by glasses, a metal matrix form, and concrete. 11 tables

  5. Characterization of radioactive waste forms. Volume 1

    International Nuclear Information System (INIS)

    Brodersen, K.; Nilsson, K.

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    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.

    2011-01-01

    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.

  7. 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.

  8. Lysimeter study of commercial reactor waste forms: waste form acquisition characterization and full-scale leaching

    International Nuclear Information System (INIS)

    1983-02-01

    This report describes work conducted at Brookhaven National Laboratory (BNL) as part of a joint program with Savannah River Laboratory. Typical full-scale (55-gallon drum size) waste forms were acquired by BNL from a boiling water reactor (BWR) and a pressurized water reactor (PWR). Liquid waste stream activity concentrations were analyzed by gamma spectroscopy. This information was used to determine the waste from activity inventory, providing the necessary source term for lysimeter and leaching experiments. Predominant radionuclides of interest include 60 Co, 137 Cs, 134 Cs, and 54 Mn. A full-scale leaching experiment was initiated by BNL encompassing four representative waste stream-solidification agent combinations. Waste streams tested include PWR evaporator concentrate (boric acid waste), BWR evaporator concentrate (sodium sulfate waste) and BWR evaporator concentrate plus ion exchange resins. Solidification agents include masonry cement, portland type III cement, and vinyl ester-styrene (Dow polymer). Analyses of leachates indicate measurable leach rates of 137 Cs, 134 Cs, and 60 Co from both BWR and PWR cement waste forms. The leach rates for both cesium isotopes in cement are at least two orders of magnitude greater than those for cobalt. Leachates from the BWR Dow polymer waste form include the same isotopes present in cement leachates, with the addition of 54 Mn. Cesium leach rates from the Dow polymer waste form are approximately one order of magnitude lower than from an equivalent cement waste form. The 60 Co cumulative fraction release, however, is approximately three times greater for the Dow polymer waste form

  9. 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

  10. High level waste fixation in cermet form

    International Nuclear Information System (INIS)

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

    1981-01-01

    Commercial and defense high level waste fixation in cermet form is being studied by personnel of the Isotopes Research Materials Laboratory, Solid State Division (ORNL). As a corollary to earlier research and development in forming high density ceramic and cermet rods, disks, and other shapes using separated isotopes, similar chemical and physical processing methods have been applied to synthetic and real waste fixation. Generally, experimental products resulting from this approach have shown physical and chemical characteristics which are deemed suitable for long-term storage, shipping, corrosive environments, high temperature environments, high waste loading, decay heat dissipation, and radiation damage. Although leach tests are not conclusive, what little comparative data are available show cermet to withstand hydrothermal conditions in water and brine solutions. The Soxhlet leach test, using radioactive cesium as a tracer, showed that leaching of cermet was about X100 less than that of 78 to 68 glass. Using essentially uncooled, untreated waste, cermet fixation was found to accommodate up to 75% waste loading and yet, because of its high thermal conductivity, a monolith of 0.6 m diameter and 3.3 m-length would have only a maximum centerline temperature of 29 K above the ambient value

  11. Impact test for solid waste forms

    International Nuclear Information System (INIS)

    Wallace, R.M.; Kelley, J.A.

    1976-03-01

    Samples of concretes and glasses being considered for incorporation of radioactive waste sludge were subjected to impact tests to determine the relationship between the energy of the impact and the resulting increase in surface area of the damaged sample. Test results indicate that the increased surface area per unit of energy input for glass waste forms is less by a factor of about three than that for concretes containing 40 wt percent simulated sludge (average values of 9.6 cm 2 /Joule and 24.7 cm 2 /Joule for glass and concrete, respectively)

  12. Waste Form Features, Events, and Processes

    International Nuclear Information System (INIS)

    R. Schreiner

    2004-01-01

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

  13. DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

    International Nuclear Information System (INIS)

    Thornton, T.A.

    2000-01-01

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

  14. Electrochemical corrosion testing of metal waste forms

    International Nuclear Information System (INIS)

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

    1999-01-01

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

    International Nuclear Information System (INIS)

    Weber, W.J.

    1981-09-01

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

  16. 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.

  17. Multibarrier waste forms. Part I. Development

    International Nuclear Information System (INIS)

    Rusin, J.M.; Lokken, R.O.; Lukacs, J.M.; Sump, K.R.; Browning, M.F.; McCarthy, G.J.

    1978-09-01

    The multibarrier concept produces a composite waste form with enhanced inertness through improvements in thermal stability, mechanical strength, and leachability by the use of coatings and metal matrices. This report describes research and development activities resulting in the demonstration of the multibarrier concept for nonradioactive simulated waste compositions. The multibarrier concept is to utilize up to three barriers to isolate radionuclides from the environment: a solid waste inner core, an impervious coating, and a metal matrix. Two inner core materials, sintered supercalcine and glass marbles, have been demonstrated. The coating barrier provides enhanced leach, impact, and oxidation resistance as well as thermal protection during encapsulation in the metal matrix. Py/Al 2 O 3 coatings deposited by chemical vapor deposition (CVD) and glass coatings have been applied to supercalcine cores to improve inertness. The purpose of the metal matrix is to improve impact resistance, protect the inner core rom any adverse environments, provide radiation shielding, and increase thermal conductivity, yielding lower internal temperatures. The development of gravity sintering and vacuum casting techniques for matrix encapsulation are discussed. Four multibarrier products were demonstrated: (1) Glass marbles encapsulated in vacuum-cast Pb-10Sn; (2) uncoated, sintered supercalcine pellets encapsulated in vacuum-cast Al-12Si; (3) glass-coated, sintered supercalcine pellets encapsulated in vacuum-cast Al-12Si; and (4) PyC/Al 2 O 3 -coated supercalcine encapsulated in gravity-sintered Cu. 23 figs., 20 tables

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

    International Nuclear Information System (INIS)

    1982-03-01

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

  19. 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

  20. Radiolytic gas production from tritiated waste forms

    International Nuclear Information System (INIS)

    Bibler, N.E.; Orebaugh, E.G.

    1977-07-01

    Radiolytic gas production during long-term storage of tritiated waste was estimated from gamma and alpha radiolysis tests to determine the extent of pressurization in sealed containers. Two forms of simulated wastes were irradiated with 60 Co gamma rays or 244 Cm alpha particles: concrete for solidification of tritiated water and vermiculite for solidification of tritiated octane or vacuum pump oil. For concrete, the gamma and alpha radiolysis results predicted that H 2 will be formed by tritium beta particles with an initial rate of 0.1 to 0.3 molecule for every 100 eV of energy absorbed. Also, as the H 2 pressure increases, this 100-eV yield decreases because of a reaction removing H 2 . Eventually, a steady state pressure that depends on the radiation intensity will be attained. For intensities less than 10 5 rads/h, the steady state pressure will be less than 20 psi. O 2 in the air sealed with the concrete will be almost completely depleted, and N 2 will be unaffected. For the organic materials sorbed onto vermiculite, the gamma and alpha radiolysis results predicted that H 2 and traces of CH 4 and CO 2 will be produced. For tritium beta particles, the 100-eV yields for H 2 based on energy sorbed by the organic materials are 4.4 for octane and 2.2 for vacuum pump oil. In the containers, steady state H 2 pressure will not be attained at pressures up to at least 200 psi. As with the concrete, O 2 will be nearly completely depleted and N 2 will be unaffected. The 100-eV yield for H 2 production was used to calculate pressure increases in conceptual tritiated waste packages. 13 figures

  1. Waste form development for a DC arc furnace

    Energy Technology Data Exchange (ETDEWEB)

    Feng, X.; Bloomer, P.E.; Chantaraprachoom, N.; Gong, M.; Lamar, D.A.

    1996-09-01

    A laboratory crucible study was conducted to develop waste forms to treat nonradioactive simulated {sup 238}Pu heterogeneous debris waste from Savannah River, metal waste from the Idaho National Engineering Laboratory (INEL), and nominal waste also from INEL using DC arc melting. The preliminary results showed that the different waste form compositions had vastly different responses for each processing effect. The reducing condition of DC arc melting had no significant effects on the durability of some waste forms while it decreased the waste form durability from 300 to 700% for other waste forms, which resulted in the failure of some TCLP tests. The right formulations of waste can benefit from devitrification and showed an increase in durability by 40%. Some formulations showed no devitrification effects while others decreased durability by 200%. Increased waste loading also affected waste form behavior, decreasing durability for one waste, increasing durability by 240% for another, and showing no effect for the third waste. All of these responses to the processing and composition variations were dictated by the fundamental glass chemistry and can be adjusted to achieve maximal waste loading, acceptable durability, and desired processing characteristics if each waste formulation is designed for the result according to the glass chemistry.

  2. 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.

  3. Review of high-level waste form properties

    International Nuclear Information System (INIS)

    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

  4. 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.

  5. Development of new waste form for treatment and disposal of concentrated liquid radioactive waste

    International Nuclear Information System (INIS)

    Kwak, Kyung Kil; Ji, Young Yong

    2010-12-01

    The radioactive waste form should be meet the waste acceptance criteria of national regulation and disposal site specification. We carried out a characterization of rad waste form, especially the characteristics of radioactivity, mechanical and physical-chemical properties in various rad waste forms. But asphalt products is not acceptable waste form at disposal site. Thus we are change the product materials. We select the development of the new process or new materials. The asphalt process is treatment of concentrated liquid and spent-resin and that we decide the Development of new waste form for treatment and disposal of concentrated liquid radioactive waste

  6. CSNF WASTE FORM DEGRADATION: SUMMARY ABSTRACTION

    International Nuclear Information System (INIS)

    CUNNANE, J.C.

    2004-01-01

    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

  7. 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.

  8. 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.

  9. Leaching behavior of phosphate-bonded ceramic waste forms

    International Nuclear Information System (INIS)

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

    1996-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Vance, E.R.; Begg, B. D.; Day, R. A.; Moricca, S.; Perera, D. S.; Stewart, M. W. A.; Carter, M. L.; McGlinn, P. J.; Smith, K. L.; Walls, P. A.; Robina, M. La

    2004-07-01

    ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and {sup 99}Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

  11. Study on the characteristics of vitrified waste form from OREOX process for dupic cycle

    International Nuclear Information System (INIS)

    Kim, J.H.; Chun, K.S.; Shin, J.M.; Kim, S.S.; Park, H.S.

    1996-01-01

    This paper describes the applicability of fly ash to the vitrification of the simulated OREOX process wastes in the aspects of productability, leachability, and preliminary characteristics of the simulated waste forms. Furthermore, vitrification of the spent filter into glass is an attractive option because it may be used as a base material for borosilicate glass. This paper also describes the possibility that the spent filter could be converted into a stable borosilicate glass form in this aspect. (author)

  12. Summary of INEL research on the iron-enriched basalt waste form

    International Nuclear Information System (INIS)

    Reimann, G.A.; Grandy, J.D.; Eddy, T.L.; Anderson, G.L.

    1992-01-01

    This report summarizes the knowledge base on the iron-enriched basalt (IEB) waste form developed at the Idaho National Engineering Laboratory (INEL) during 1979--1982. The results presented discuss the applicability of IEB in converting retrieved transuranic (TRU) waste from INEL's Radioactive Waste Management Complex (RWMC) into a vitreous/ceramic (glassy/rock) stable waste form suitable for permanent disposal in an appropriate repository, such as the Waste Isolation Pilot Plant (WIPP) in New Mexico. Borosilicate glass (BSG), the approved high-level waste form, appears unsuited for this application. Melting the average waste-soil mix from the RWMC produces the IEB composition and attempting to convert IEB to the BSG composition would require additions of substantial B 2 0 3 , Na, and SiO 2 (glass frit). IEB requires processing temperatures of 1400 to 1600 degrees C, depending upon the waste composition. Production of the IEB waste form, using Joule heated melters, has proved difficult in the past because of electrode and refractory corrosion problems associated with the high temperature melts. Higher temperature electric melters (arc and plasma) are available to produce this final waste form. Past research focused on extensive slag property measurements, waste form leachability tests, mechanical, composition, and microstructure evaluations, as well as a host of experiments to improve production of the waste form. Past INEL studies indicated that the IEB glass-ceramic is a material that will accommodate and stabilize a wide range of heterogeneous waste materials, including long lived radionuclides and scrap metals, while maintaining a superior level of chemical and physical performance characteristics. Controlled cooling of the molten IEB and subsequent heat treatment will produce a glass-ceramic waste form with superior leach resistance

  13. The construction of solid waste form test facility

    International Nuclear Information System (INIS)

    Park, Hun Hwee; Kim, Joon Hyung; Lee, Byung Jik; Koo, Jun Mo; Kim, Jeong Guk; Jung, In Ha

    1990-03-01

    The solid waste form test facility (SWFTF) to test and/or evaluate the characteristics of waste forms, such as homogeniety, mechanical properties, thermal properties, waste resistance and leachability, have been constructed, and some equipments for testing actual waste forms has been purchased; radiocative monitoring system, glove box for the manipulator repair room, and uninteruppted power supply system, et al. Classifications of radioactive wastes, basic requirements and criteria to be considered during waste management were also reviewed. Some of the described items above have been standardized for the purpose of indigenigation. Therefore, safety assurance of waste forms, as well as increase in the range of participating of domestic companies in construction of further nuclear facilities could be obtained as results through constructing this facility. In the furture this facility is going to be utilized not only for the inspection of waste forms but also for the periodic decontamination for extending the life time of some expensive radiological equipments using remote handling techniques. (author)

  14. 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.

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

    International Nuclear Information System (INIS)

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

    1981-09-01

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

  16. Waste acceptance product specifications for vitrified high-level waste forms

    International Nuclear Information System (INIS)

    Applewhite-Ramsey, A.; Sproull, J.F.

    1993-01-01

    The Nuclear Waste Policy Act of 1982 mandated that all high-level waste (HLW) be sent to a federal geologic repository for permanent disposal. DOE published the Environmental Assessment in 1982 which identified borosilicate glass as the chosen HLW form. 1 In 1985 the Department of Energy instituted a Waste Acceptance Process to assure that DWPF glass waste forms would be acceptable to such a repository. This assurance was important since production of waste forms will precede repository construction and licensing. As part of this Waste Acceptance Process, the DOE Office of Civilian Radioactive Waste Management (RW) formed the Waste Acceptance Committee (WAC). The WAC included representatives from the candidate repository sites, the waste producing sites and DOE. The WAC was responsible for developing the Waste Acceptance Preliminary Specifications (WAPS) which defined the requirements the waste forms must meet to be compatible with the candidate repository geologies

  17. Influence of system considerations on waste form design

    International Nuclear Information System (INIS)

    Bauer, A.A.; Matthews, S.C.; Peterson, R.W.

    1979-01-01

    The design of waste forms is constrained by waste management system considerations imposed during generation, treatment, packaging, transportation, storage, and isolation. In the isolation phase, the waste form provides one of the barriers to release in a multibarrier system that includes the natural geologic and hydrologic barriers as well as other engineered barriers

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

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  19. PASSIVATION LAYER STABILITY OF A METALLIC ALLOY WASTE FORM

    International Nuclear Information System (INIS)

    Williamson, M.; Mickalonis, J.; Fisher, D.; Sindelar, R.

    2010-01-01

    Alloy waste form development under the Waste Forms Campaign of the DOE-NE Fuel Cycle Research and Development program includes the process development and characterization of an alloy system to incorporate metal species from the waste streams generated during nuclear fuel recycling. This report describes the tests and results from the FY10 activities to further investigate an Fe-based waste form that uses 300-series stainless steel as the base alloy in an induction furnace melt process to incorporate the waste species from a closed nuclear fuel recycle separations scheme. This report is focused on the initial activities to investigate the formation of oxyhydroxide layer(s) that would be expected to develop on the Fe-based waste form as it corrodes under aqueous repository conditions. Corrosion tests were used to evaluate the stability of the layer(s) that can act as a passivation layer against further corrosion and would affect waste form durability in a disposal environment.

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

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.S.; Cunnane, J.; Sutaria, M.; Kurokawa, S.; Mayberry, J.

    1994-04-01

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

  1. Alternate nuclear waste forms and interactions in geologic media

    International Nuclear Information System (INIS)

    Boatner, L.A.; Battle, G.C. Jr.

    1981-04-01

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

  2. The DWPF waste form qualification program

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  3. Statistical criteria to evaluate homogeneity of radioactive waste forms

    International Nuclear Information System (INIS)

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

    1987-08-01

    A set of statistical tests is proposed to evaluate the homogeneity of radioactive waste forms. These criteria were applied to cemented waste forms by using natural thorium and uranium compounds as tracers to evaluate the chemical homogeneity of the final product. The mixing technique used to produce the cemented waste forms proved to be good enough as a standard method to obtain a homogeneous product. (Author) [pt

  4. Full-scale leaching study of commercial reactor waste forms

    International Nuclear Information System (INIS)

    Kalb, P.D.; Colombo, P.

    1984-01-01

    This paper describes a full-scale leaching experiment which has been conducted at Brookhaven National Laboratory (BNL) to study the release of radionuclides from actual commercial reactor waste forms. While many studies characterizing the leaching behavior of simulated laboratory-scale waste forms have been performed, this program represents one of the first attempts in the United States to quantify activity releases for real, full-scale waste forms. 5 references, 5 figures, 1 table

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

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Benjamin D., E-mail: Benjamin.Williams@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-54, Richland, WA 99352 (United States); Neeway, James J., E-mail: James.Neeway@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-54, Richland, WA 99352 (United States); Snyder, Michelle M.V., E-mail: Michelle.ValentaSnyder@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-54, Richland, WA 99352 (United States); Bowden, Mark E., E-mail: Mark.Bowden@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-54, Richland, WA 99352 (United States); Amonette, James E., E-mail: Jim.Amonette@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-54, Richland, WA 99352 (United States); Arey, Bruce W., E-mail: Bruce.Arey@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-54, Richland, WA 99352 (United States); Pierce, Eric M., E-mail: pierceem@ornl.gov [Oak Ridge National Laboratory, PO Box 2008, MS-6035, Room 372, Oak Ridge, TN 37831 (United States); Brown, Christopher F., E-mail: Christopher.Brown@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-54, Richland, WA 99352 (United States); Qafoku, Nikolla P., E-mail: Nik.Qafoku@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-54, Richland, WA 99352 (United States)

    2016-05-15

    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. - Highlights: • Simulated Hanford waste was treated by the Fluidized Bed Steam Reformer (FBSR) process. • The FBSR granular product was encapsulated in a geopolymer monolith. • Leach tests were performed to examine waste form performance. • XRD revealed the structure of a previously unreported sodium aluminosilicate phase. • Monolithing of granular waste forms may lead to a reduction in crystallinity.

  6. DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

    International Nuclear Information System (INIS)

    CUNNANE, J.

    2004-01-01

    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 ''upper-limit'' (i

  7. Testing waste forms containing high radionuclide loadings

    International Nuclear Information System (INIS)

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

    1986-01-01

    The Low-Level Waste Data Base Development - EPICOR-II Resin/Liner Investigation Program of the US Nuclear Regulatory Commission (NRC) is obtaining information on radioactive waste during NRC-prescribed tests and in a disposal environment. This paper describes the resin solidification task of that program, including the present status and results to date

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

    International Nuclear Information System (INIS)

    Finkeldei, Sarah Charlotte

    2015-01-01

    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 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 2 based pyrochlores. ZrO 2 - Nd 2 O 3 pellets with pyrochlore and defect

  9. 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.

  10. 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.

  11. 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

  12. Compatibility testing of vitrified waste forms

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1978-01-01

    The compatibility of vitrified radioactive waste with candidate canister materials will be evaluated with both cast and in-can melted vitrified waste. Both real and simulated sludges will be used. In addition, the compatibility of these materials with salt from a possible final storage location will be determined. Cast vitrified waste will be tested with ASTM A 333 and ASTM A 516 low-carbon steels and Type 304L stainless steel at 100, 600 and 800 0 C. Cast vitrified waste that has been devitrified by heat treatment will be tested at 100 0 C. Two types of test specimens will be used with either simulated or real sludges: (1) unsealed capsules made of pieces of mill-finished pipe into which vitrified waste is cast, and (2) sealed capsules containing a small container of vitrified waste identical to the ones in the unsealed capsule. In-can melted vitrified waste will be tested with synthetic sludge only and with ASTM A 333 and ASTM A 516 low-carbon steels, Type 304L stainless steel and Inconel 600. Two types of tests will be carried out: (1) melting vitrified waste in miniature metal canisters and (2) exposure of small (carefully measured) metal coupons to molten glass. The air oxidation rates of candidate canister materials will be determined, and specimens will also be exposed to salt from Drill Hole AEC-8 in Carlsbad, New Mexico. Sealed capsules containing an ASTM A 516 low-carbon steel or Type 304L stainless steel specimen partially embedded in a small block of salt will be heated

  13. Leaching studies of low-level radioactive waste forms

    International Nuclear Information System (INIS)

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

    1985-01-01

    A research program has been underway at the Brookhaven National Laboratory to investigate the release of radionuclides from low-level waste forms under laboratory conditions. This paper describes the leaching behavior of Cs-137 from two major low-level waste streams, that is, ion exchange bead resin and boric acid concentrate, solidified in Portland cement. The resultant leach data are employed to evaluate and predict the release behavior of Cs-137 from low-level waste forms under field burial conditions

  14. Radiation transport in high-level waste form

    International Nuclear Information System (INIS)

    Arakali, V.S.; Barnes, S.M.

    1992-01-01

    The waste form selected for vitrifying high-level nuclear waste stored in underground tanks at West Valley, NY is borosilicate glass. The maximum radiation level at the surface of a canister filled with the high-level waste form is prescribed by repository design criteria for handling and disposition of the vitrified waste. This paper presents an evaluation of the radiation transport characteristics for the vitreous waste form expected to be produced at West Valley and the resulting neutron and gamma dose rates. The maximum gamma and neutron dose rates are estimated to be less than 7500 R/h and 10 mRem/h respectively at the surface of a West Valley canister filled with borosilicate waste glass

  15. 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.

  16. Results of field testing of waste forms using lysimeters

    International Nuclear Information System (INIS)

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

    1990-01-01

    The TMI-2 [Three Mile Island Unit 2] EPICOR-II Resin/Liner Low-Level Waste Data Base Development Program, funded by the U.S. Nuclear Regulatory Commission (NRC), is obtaining information on the performance of radioactive waste in a disposal environment. Waste forms fabricated using ion-exchange resins from EPICOR-II prefilters employed in the cleanup of the Three Mile Island Nuclear Power Station are being tested to: (a) develop a low-level waste data base, and (b) obtain information on survivability of waste forms in a disposal environment. This paper updates field testing of those waste forms during FY-1989. The results of the lysimeters is presented and the use of lysimeter data in performance assessment is discussed. 16 refs., 10 figs., 5 tabs

  17. Development and evaluation of candidate high-level waste forms

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.

    1981-01-01

    Some seventeen candidate waste forms have been investigated under US Department of Energy programs as potential media for the immobilization and geologic disposal of the high-level radioactive wastes (HLW) resulting from chemical processing of nuclear reactor fuels and targets. Two of these HLW forms were selected at the end of fiscal year (FY) 1981 for intensive development if FY 1982 to 1983. Borosilicate glass was continued as the reference form. A crystalline ceramic waste form, SYNROC, was selected for further product formulation and process development as the alternative to borosilicate glass. This paper describes the bases on which this decision was made

  18. Testing and evaluation of polyethylene and sulfur cement waste forms

    International Nuclear Information System (INIS)

    Franz, E.M.; Kalb, P.D.; Colombo, P.

    1985-01-01

    This paper discusses the results of recent studies related to the use of polyethylene and modified sulfur cement as new binder materials for the improved solidification of low-level wastes. Waste streams selected for this study include those which result from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those that remain problematic for solidification using contemporary agents (ion-exchange resins). Maximum waste loadings were determined for each waste type. Recommended waste loadings of 70 wt % sodium sulfate, 50 wt % boric acid, 40 wt % incinerator ash and 30 wt % ion exchange resins, which are based on process control and waste form performance considerations are reported for polyethylene. For sulfur cement the recommended waste loadings of 40 wt % sodium sulfate and boric acid salts and 43 wt % incinerator ash are reported. However, incorporation of ion-exchange resin waste in modified sulfur cement is not recommended due to poor waste form performance. The work presented in this paper will, in part, present data that can be used to assess the acceptability of polyethylene and modified sulfur cement waste forms to meet the requirements of 10 CFR 61. 8 refs., 10 figs., 6 tabs

  19. 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.

  20. NNWSI waste form testing at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Bates, J.K.; Gerding, T.J.; Abrajano, T.A. Jr.; Ebert, W.L.; Mazer, J.J.

    1988-11-01

    The Nevada Nuclear Waste Storage Investigation (NNWSI) Project is investigating the tuff beds of Yucca Mountain, Nevada, as a potential location for a high-level radioactive waste repository. As part of the waste package development portion of this project, experiments are being performed by the Chemical Technology Division of Argonne National Laboratory to study the behavior of the waste form under anticipated repository conditions. These experiments include the development and performance of a test to measure waste form behavior in unsaturated conditions and the performance of experiments designed to study the behavior of waste package components in an irradiated environment. Previous reports document developments in these areas through 1986. This report summarizes progress during the period January--June 1987, 19 refs., 17 figs., 20 tabs

  1. Broken Homes: Stable Risk, Changing Reasons, Changing Forms.

    Science.gov (United States)

    Sweetser, Dorrian Apple

    1985-01-01

    Cohort membership and two measures of social disadvantage were used as explanatory variables in analysis of the risk of growing up in a broken home and of the living arrangements of children with broken homes. The risk of a broken home by age 16 proved to be stable across cohorts and greater for those from disadvantaged homes. (Author/BL)

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

    International Nuclear Information System (INIS)

    1979-01-01

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

  3. Disposal criticality analysis methodology for fissile waste forms

    International Nuclear Information System (INIS)

    Davis, J.W.; Gottlieb, P.

    1998-03-01

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

  4. Naturally occurring glasses: analogues for radioactive waste forms

    International Nuclear Information System (INIS)

    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 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

  5. Testing waste forms containing high radionuclide loadings

    International Nuclear Information System (INIS)

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

    1986-01-01

    The Low-Level Waste Data Base Development - EPICOR-II Resin/Liner Investigation Program funded by the US Nuclear Regulatory Commission (NRC) is obtaining information on radioactive waste during NRC-prescribed tests and in a disposal environment. This paper describes the resin solidification task of that program, including the present status and results to date. An unusual aspect of this investigation is the use of commercial grade, ion exchange resins that have been loaded with over five times the radioactivity normally seen in a commercial application. That dramatically increases the total radiation dose to the resins. The objective of the resin solidification task is to determine the adequacy of test procedures specified by NRC for ion exchange resins having high radionuclide loadings

  6. 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

  7. Performance assessment of solidified tritiated waste forms

    International Nuclear Information System (INIS)

    Deneanu, N.; Antonescu, M.

    1997-01-01

    Present research continues an investigation of use of immobilization materials (clay, natural zeolites and cement binders with appropriate additive sand, volcanic tuff) for immobilization of the tritiated liquid wastes. Laboratory results obtained to date demonstrate that the leachate from the wastes passes acceptance criteria for radioactivity. The tritium release rates were 10 -4 to 8 x 10 -5 g / cm 2 · day. The use of coating materials to reduce the leaching of tritium from tritiated cement blocks has been studied. Paraffin and bitumen coatings were made by dipping the cement blocks into the coating, whereas other materials were brushed on the surface. The leaching rates of tritium were around 6 x 10 -4 g/cm 2 · day. Since encapsulation can be regarded as inside in-situ method, the use of high density polyethylene as a container was investigated as an alternative. Of the types of absorbents evaluated, volcanic tuff was found to be very effective at absorbing tritiated liquid waste. The leaching rates of tritium from tritiated liquids into a 0.1 cm thick polyethylene container were 10 -6 g/ cm 2 · day. (authors)

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  9. Fundamental Aspects of Zeolite Waste Form Production by Hot Isostatic Pressing

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jordan, Jacob A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-02-01

    The direct conversion of iodine-bearing sorbents into a stable waste form is a research topic of interest to the US Department of Energy. The removal of volatile radioactive 129I from the off-gas of a nuclear fuel reprocessing facility will be necessary in order to comply with the regulatory requirements that apply to facilities sited within the United States (Jubin et al., 2012a), and any iodine-containing media or solid sorbents generated by this process would contain 129I and would be destined for eventual geological disposal. While recovery of iodine from some sorbents is possible, a method to directly convert iodineloaded sorbents to a durable waste form with little or no additional waste materials being formed and a potentially reduced volume would be beneficial. To this end, recent studies have investigated the conversion of iodine-loaded silver mordenite (I-AgZ) directly to a waste form by hot isostatic pressing (HIPing) (Bruffey and Jubin, 2015). Silver mordenite (AgZ), of the zeolite class of minerals, is under consideration for use in adsorbing iodine from nuclear reprocessing off-gas streams. Direct conversion of I-AgZ by HIPing may provide the following benefits: (1) a waste form of high density that is tolerant to high temperatures, (2) a waste form that is not significantly chemically hazardous, and (3) a robust conversion process that requires no pretreatment.

  10. Equipping a glovebox for waste form testing and characterization of plutonium bearing materials

    International Nuclear Information System (INIS)

    Noy, M.; Johnson, S.G.; Moschetti, T.L.

    1997-01-01

    The recent decision by the Department of Energy to pursue a hybrid option for the disposition of weapons plutonium has created the need for additional facilities that can examine and characterize waste forms that contain Pu. This hybrid option consists of the placement of plutonium into stable waste forms and also into mixed oxide fuel for commercial reactors. Glass and glass-ceramic waste forms have a long history of being effective hosts for containing radionuclides, including plutonium. The types of tests necessary to characterize the performance of candidate waste forms include: static leaching experiments on both monolithic and crushed waste forms, microscopic examination, and density determination. Frequently, the respective candidate waste forms must first be produced using elevated temperatures and/or high pressures. The desired operations in the glovebox include, but are not limited to the following: (1) production of vitrified/sintered samples, (2) sampling of glass from crucibles or other vessels, (3) preparing samples for microscopic inspection and monolithic and crushed static leach tests, and (4) performing and analyzing leach tests in situ. This paper will describe the essential equipment and modifications that are necessary to successfully accomplish the goal of outfitting a glovebox for these functions

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    Highlights: • Decommissioning concrete waste recycling and disposal. • Compressive strength of cement waste form. • Characteristic of thermal resistance and leaching of cement waste form. - Abstract: In Korea, the decontamination and decommissioning of KRR-1, 2 at KAERI have been under way. The decommissioning of the KRR-2 was finished completely by 2011, whereas the decommissioning of KRR-1 is currently underway. A large quantity of slightly contaminated concrete waste has been generated from the decommissioning projects. The concrete wastes, 83ea of 200 L drums, and 41ea of 4 m 3 containers, were generated in the decommissioning projects. The conditioning of concrete waste is needed for final disposal. Concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled with a void space after concrete rubble pre-placement into 200 L drums. Thus, this research developed an optimizing mixing ratio of concrete waste, water, and cement, and evaluated the characteristics of a cement waste form to meet the requirements specified in the disposal site specific waste acceptance criteria. The results obtained from a compressive strength test, leaching test, and thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested as an optimized mixing ratio of 75:15:10. In addition, the compressive strength of the cement waste form was satisfied, including a fine powder up to a maximum of 40 wt% in concrete debris waste of about 75%. According to the scale-up test, the mixing ratio of concrete waste, water, and cement is 75:10:15, which meets the satisfied compressive strength because of an increase in the particle size in the waste

  13. Method for forming microspheres for encapsulation of nuclear waste

    Science.gov (United States)

    Angelini, Peter; Caputo, Anthony J.; Hutchens, Richard E.; Lackey, Walter J.; Stinton, David P.

    1984-01-01

    Microspheres for nuclear waste storage are formed by gelling droplets containing the waste in a gelation fluid, transferring the gelled droplets to a furnace without the washing step previously used, and heating the unwashed gelled droplets in the furnace under temperature or humidity conditions that result in a substantially linear rate of removal of volatile components therefrom.

  14. A U-Bearing Composite Waste Form for Electrochemical Processing Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Chen, X; Ebert, W. L.; Indacochea, J.E.

    2018-04-01

    Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phases that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases. (c) 2018 Elsevier B.V. All rights reserved.

  15. Defining a metal-based waste form for IFR pyroprocessing wastes

    International Nuclear Information System (INIS)

    McDeavitt, S.M.; Park, J.Y.; Ackerman, J.P.

    1994-01-01

    Pyrochemical electrorefining to recover actinides from metal nuclear fuel is a key element of the Integral Fast Reactor (IFR) fuel cycle. The process separates the radioactive fission products from the long-lived actinides in a molten LiCl-KCl salt, and it generates a lower waste volume with significantly less long-term toxicity as compared to spent nuclear fuel. The process waste forms include a mineral-based waste form that will contain fission products removed from an electrolyte salt and a metal-based waste form that will contain metallic fission products and the fuel cladding and process materials. Two concepts for the metal-based waste form are being investigated: (1) encapsulating the metal constituents in a Cu-Al alloy and (2) alloying the metal constituents into a uniform stainless steel-based waste form. Results are given from our recent studies of these two concepts

  16. A U-bearing composite waste form for electrochemical processing wastes

    Science.gov (United States)

    Chen, X.; Ebert, W. L.; Indacochea, J. E.

    2018-04-01

    Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phases that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases.

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

    International Nuclear Information System (INIS)

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

    1983-09-01

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

  18. Evaluation of conditioned high-level waste forms

    International Nuclear Information System (INIS)

    Mendel, J.E.; Turcotte, R.P.; Chikalla, T.D.; Hench, L.L.

    1983-01-01

    The evaluation of conditioned high-level waste forms requires an understanding of radiation and thermal effects, mechanical properties, volatility, and chemical durability. As a result of nuclear waste research and development programs in many countries, a good understanding of these factors is available for borosilicate glass containing high-level waste. The IAEA through its coordinated research program has contributed to this understanding. Methods used in the evaluation of conditioned high-level waste forms are reviewed. In the US, this evaluation has been facilitated by the definition of standard test methods by the Materials Characterization Center (MCC), which was established by the Department of Energy (DOE) in 1979. The DOE has also established a 20-member Materials Review Board to peer-review the activities of the MCC. In addition to comparing waste forms, testing must be done to evaluate the behavior of waste forms in geologic repositories. Such testing is complex; accelerated tests are required to predict expected behavior for thousands of years. The tests must be multicomponent tests to ensure that all potential interactions between waste form, canister/overpack and corrosion products, backfill, intruding ground water and the repository rock, are accounted for. An overview of the status of such multicomponent testing is presented

  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. Fundamental Thermodynamics of Actinide-Bearing Mineral Waste Forms

    International Nuclear Information System (INIS)

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

    1999-01-01

    The recent arms reduction treaties between the U.S. and Russia have resulted in inventories of plutonium in excess of current defense needs. Storage of this material poses significant, and unnecessary, risks of diversion, especially for Russia whose infrastructure for protecting these materials has been weakened since the collapse of the Soviet Union. Moreover, maintaining and protecting these materials in their current form is costly. The United States has about sixty metric tons of excess plutonium, half of which is high-purity weapon material. This high purity material will be converted into mixed oxide (MOX) fuel for use in nuclear reactors. The less pure excess plutonium does not meet the specifications for MOX fuel and will not be purified to meet the fuel specifications. Instead, it will be immobilized directly in a ceramic. The ceramic will be encased in a high level waste (HLW) glass monolith (i.e., the can-in-canister option) thus making a form that simulates the intrinsic security of spent nuclear fuel. The immobilized product will be placed in a HLW repository. To meet the repository requirements, the product must be shown to be durable for the intended storage time, the host matrix must be stable in the radiation environment, the solubility and leaching characteristics of the plutonium in the host material must be established, and optimum processing parameters must be determined for the entire compositional envelope of feed materials. 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 forms proposed as immobilization matrices. However, the relevant thermodynamic data (e.g., enthalpy, entropy, and heat capacity) for the ceramic forms are severely lacking and this information gap directly affects the Energy Department's ability to license the disposal matrices and methods. High-temperature solution

  1. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

    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. 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.

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

    International Nuclear Information System (INIS)

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

    1985-09-01

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

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

    International Nuclear Information System (INIS)

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

    1982-09-01

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

  5. Low-level radioactive waste form qualification testing

    International Nuclear Information System (INIS)

    Sohal, M.S.; Akers, D.W.

    1998-06-01

    This report summarizes activities that have already been completed as well as yet to be performed by the Idaho National Engineering and Environmental Laboratory (INEEL) to develop a plan to quantify the behavior of radioactive low-level waste forms. It briefly describes the status of various tasks, including DOE approval of the proposed work, several regulatory and environmental related documents, tests to qualify the waste form, preliminary schedule, and approximate cost. It is anticipated that INEEL and Brookhaven National Laboratory will perform the majority of the tests. For some tests, services of other testing organizations may be used. It should take approximately nine months to provide the final report on the results of tests on a waste form prepared for qualification. It is anticipated that the overall cost of the waste quantifying service is approximately $150,000. The following tests are planned: compression, thermal cycling, irradiation, biodegradation, leaching, immersion, free-standing liquid tests, and full-scale testing

  6. Alternative waste form development - low-temperature pyrolytic carbon coatings

    International Nuclear Information System (INIS)

    Oma, K.H.; Rusin, J.M.; Kidd, R.W.; Browning, M.F.

    1981-01-01

    Although several chemical vapor deposition (CVD) - coated waste forms have been successfully produced, some major disadvantages associated with the high-temperature fluidized-bed CVD coating process exist. To overcome these disadvantages, the Pacific Northwest Laboratory has initiated the development of a pyrolytic carbon CVD coating system to coat large waste-form particles at temperatures ranging from 400 to 500/degree/C. This relatively simple system has been used to coat kilogram quantities of simulated waste-glass marbles. Further development of this system could result in a viable process to coat bulk quantities of both glass and ceramic waste forms. This paper discusses various aspects of the development work, including coating techniques, parametric study, and coater equipment. 10 refs

  7. Low-level radioactive waste form qualification testing

    Energy Technology Data Exchange (ETDEWEB)

    Sohal, M.S.; Akers, D.W.

    1998-06-01

    This report summarizes activities that have already been completed as well as yet to be performed by the Idaho National Engineering and Environmental Laboratory (INEEL) to develop a plan to quantify the behavior of radioactive low-level waste forms. It briefly describes the status of various tasks, including DOE approval of the proposed work, several regulatory and environmental related documents, tests to qualify the waste form, preliminary schedule, and approximate cost. It is anticipated that INEEL and Brookhaven National Laboratory will perform the majority of the tests. For some tests, services of other testing organizations may be used. It should take approximately nine months to provide the final report on the results of tests on a waste form prepared for qualification. It is anticipated that the overall cost of the waste quantifying service is approximately $150,000. The following tests are planned: compression, thermal cycling, irradiation, biodegradation, leaching, immersion, free-standing liquid tests, and full-scale testing.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. In addition, 83 drums of 200 l, and 41 containers of 4 m 3 of concrete waste were generated. Conditioning of concrete waste is needed for final disposal. Concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled into a void space after concrete rubble pre-placement into 200 l drums. Thus, this research developed an optimizing mixing ratio of concrete waste, water, and cement, and evaluated the characteristics of a cement waste form to meet the requirements specified in the disposal site specific waste acceptance criteria. The results obtained from compressive strength test, leaching test, and thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10 as the optimized mixing ratio. In addition, the compressive strength of cement waste form was satisfied, including fine powder up to a maximum 40 wt% in concrete debris waste of about 75%. (authors)

  9. 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

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

    International Nuclear Information System (INIS)

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

    1993-11-01

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

  11. Cemented Solidified Biomass Waste Form Under Drastic Climatic Conditions

    International Nuclear Information System (INIS)

    Bayoumi, T.A.

    2013-01-01

    Immobilizing radioactive waste generated during different nuclear applications is a matter of great importance for safe disposal. Waste forms buried in a shallow burial disposal facility as low level and intermediate level radioactive wastes could be solidified by cement or other inert material that characterized by adequate strength against drastic climatic changes. In this study, a biomass generated from bioaccumulation of aquatic plants containing hazardous radioactive nuclides were dried, grinded, stabilized by cement that cured for 28 days then subjected to qualification during different aggressive conditions. The produced solidified waste form was followed by detecting mechanical strength and porosity measurements as well as spectroscopic analysis using infrared and X-ray. Fractured waste form sample was investigated using scanning electron microscope to examine the effect of different undesirable climatic events during extending disposal durations. It was found that the cement solidification of the aquatic plants generated from the phytoremediation of radioactive simulate is an essential step before disposal. The obtained cemented waste forms have adequate resistance under extreme climatic conditions

  12. 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.

  13. Palmitic acid/polypyrrole composites as form-stable phase change materials for thermal energy storage

    International Nuclear Information System (INIS)

    Silakhori, Mahyar; Metselaar, Hendrik Simon Cornelis; Mahlia, Teuku Meurah Indra; Fauzi, Hadi; Baradaran, Saeid; Naghavi, Mohammad Sajad

    2014-01-01

    Highlights: • A novel phase change composite of palmitic acid–polypyrrole(PA–PPy) was fabricated. • Thermal properties of PA–PPy are characterized in different mass ratios of PA–PPy. • Thermal cycling test showed that form stable PCM had a favorable thermal reliability. - Abstract: In this study a novel palmitic acid (PA)/polypyrrole (PPy) form-stable PCMs were readily prepared by in situ polymerization method. PA was used as thermal energy storage material and PPy was operated as supporting material. Form-stable PCMs were investigated by SEM (scanning electron microscopy) and FTIR (Fourier transform infrared spectrometer) analysis that illustrated PA Particles were wrapped by PPy particles. XRD (X-ray diffractometer) was used for crystalline phase of PA/PPy composites. Thermogravimetry analysis (TGA) and differential scanning calorimetry (DSC) were used for investigating Thermal stability and thermal energy storage properties of prepared form-stable PCMs. According to the obtained results the form stable PCMs exhibited favorable thermal stability in terms of their phase change temperature. The form-stable PCMs (79.9 wt% loading of PA) were considered as the highest loading PCM with desirable latent heat storage of 166.3 J/g and good thermal stability. Accelerated thermal cycling tests also showed that form stable PCM had an acceptable thermal reliability. As a consequence of acceptable thermal properties, thermal stability and chemical stability, we can consider the new kind of form stable PCMs for low temperature solar thermal energy storage applications

  14. Hydrated phases and pore solution composition in cementsolidified saltstone waste forms

    Directory of Open Access Journals (Sweden)

    Philip J.

    2013-07-01

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

  15. NNWSI waste form test method for unsaturated disposal conditions

    International Nuclear Information System (INIS)

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

    1985-03-01

    A test method has been developed to measure the release of radionuclides from the waste package under simulated NNWSI repository conditions, and to provide information concerning materials interactions that may occur in the repository. Data are presented from Unsaturated testing of simulated Savannah River Laboratory 165 glass completed through 26 weeks. The relationship between these results and those from parametric and analog testing are described. The data indicate that the waste form test is capable of producing consistent, reproducible results that will be useful in evaluating the role of the waste package in the long-term performance of the repository. 6 refs., 7 figs., 5 tabs

  16. Pirm wastes: permanent isolation in rock-forming minerals

    International Nuclear Information System (INIS)

    Smyth, J.R.; Vidale, R.J.; Charles, R.W.

    1977-01-01

    The most practical system for permanent isolation of radioactive wastes in granitic and pelitic environments may be one which specifically tailors the waste form to the environment. This is true because if recrystallization of the waste form takes place within the half-lives of the hazardous radionuclides, it is likely to be the rate-controlling step for release of these nuclides to the ground-water system. The object of the proposed waste-form research at Los Alamos Scintific Laboratory (LASL) is to define a phase assemblage which will minimize chemical reaction with natural fluids in a granitic or pelitic environment. All natural granites contain trace amounts of all fission product elements (except Tc) and many contain minor amounts of these elements as major components of certain accessory phases. Observation of the geochemistry of fission-product elements has led to the identification of the natural minerals as target phases for research. A proposal is made to experimentally determine the amounts of fission product elements which can stably be incorporated into the phases listed below and to determine the leachability of the assemblage this produced using fluids typical of the proposed environments at the Nevada Test Site. This approach to waste isolation satisfies the following requirements: (1) It minimizes chemical reaction with the environment (i.e., recrystallization) which is likely to be the rate-controlling step for release of radionuclides to groundwater; (2) Waste loading (hence temperature) can be easily varied by dilution with material mined from the disposal site; (3) No physical container is required; (4) No maintenance is required (permanent); (5) The environment acts as a containment buffer. It is proposed that such wastes be termed PIRM wastes, for Permanent Isolation in Rock-forming Minerals

  17. Consolidated waste forms: glass marbles and ceramic pellets

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  18. 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.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  20. Transmutation of stable isotopes and deactivation of radioactive waste in growing biological systems

    International Nuclear Information System (INIS)

    Vysotskii, Vladimir I.; Kornilova, Alla A.

    2013-01-01

    Highlights: ► The phenomena of isotope transmutation in growing microbiological cultures were investigated. ► Transmutation in microbiological associations is 20 times more effective than in pure cultures. ► Transmutation of radioactive nuclei to stable isotopes in such associations was investigated. ► The most accelerated rate of Cs 137 to stable Ba 138 isotope transmutation was 310 days. ► “Microbiological deactivation” may be used for deactivation of Chernobyl and Fukushima areas. - Abstract: The report presents the results of qualifying examinations of stable and radioactive isotopes transmutation processes in growing microbiological cultures. It is shown that transmutation of stable isotopes during the process of growth of microbiological cultures, at optimal conditions in microbiological associations, is 20 times more effective than the same transmutation process in the form of “one-line” (pure) microbiological cultures. In the work, the process of direct, controlled decontamination of highly active intermediate lifetime and long-lived reactor isotopes (reactor waste) through the process of growing microbiological associations has been studied. In the control experiment (flask with active water but without microbiological associations), the “usual” law of nuclear decay applies, and the life-time of Cs 137 isotope was about 30 years. The most rapidly increasing decay rate, which occurred with a lifetime τ * ≈ 310 days (involving an increase in rate, and decrease in lifetime by a factor of 35 times) was observed in the presence of Ca salt in closed flask with active water contained Cs 137 solution and optimal microbiological association

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

    International Nuclear Information System (INIS)

    Akers, D.W.; Kraft, N.C.; Mandler, J.W.

    1994-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

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

    International Nuclear Information System (INIS)

    Owen, D.E.

    1981-08-01

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

  4. 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.

  5. Performance testing of waste forms in a tuff environment

    International Nuclear Information System (INIS)

    Oversby, V.M.

    1983-11-01

    This paper describes experimental work conducted to establish the chemical composition of water which will have reacted with Topopah Spring Member tuff prior to contact with waste packages. The experimental program to determine the behavior of spent fuel and borosilicate glass in the presence of this water is then described. Preliminary results of experiments using spent fuel segments with defects in the Zircaloy cladding are presented. Some results from parametric testing of a borosilicate glass with tuff and 304L stainless steel are also discussed. Experiments conducted using Topopah Spring tuff and J-13 well water have been conducted to provide an estimate of the post-emplacement environment for waste packages in a repository at Yucca Mountain. The results show that emplacement of waste packages should cause only small changes in the water chemistry and rock mineralogy. The changes in environment should not have any detrimental effects on the performance of metal barriers or waste forms. The NNWSI waste form testing program has provided preliminary results related to the release rate of radionuclides from the waste package. Those results indicate that release rates from both spent fuel and borosilicate glass should be below 1 part in 10 5 per year. Future testing will be directed toward making release rate testing more closely relevant to site specific conditions. 17 references, 7 figures

  6. Transuranic waste form characterization and data base. Executive summary

    International Nuclear Information System (INIS)

    1980-01-01

    The Transuranic Waste Form Characterization and Data Base (Volume 1) provides a wide range of information from which a comprehensive data base can be established and from which standards and criteria can be developed for the present NRC waste management program. Supplementary information on each of the areas discussed in Volume 1 is presented in Appendices A through K (Volumes 2 and 3). The structure of the study (Volume 1) is outlined and appendices of Volumes 2 and 3 correlate with each main section of the report. The Executive Summary reviews the sources, quantities, characteristics and treatment of transuranic wastes in the United States. Due to the variety of potential treatment processes for transuranic wastes, the end products for long-term storage may have corresponding variations in quantities and characteristics

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

    International Nuclear Information System (INIS)

    Oversby, V.M.

    1994-05-01

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

  8. The effects of radiation on intermediate-level waste forms. Task 3 characterization of radioactive waste forms a series of final reports (1985-89) no. 10

    International Nuclear Information System (INIS)

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

    1991-01-01

    The purpose of this programme was to determine the effects of radiation on the properties of intermediate-level waste forms relevant to their storage and disposal. It had two overall aims: to provide immediate data on the effect of radiation on important European ILW waste forms through accelerated laboratory tests; and to develop an understanding of the degradation processes so that long-term, low dose rate effects can be predicted with confidence from short-term, high dose rate experiments. The programme included cement waste forms containing inorganic wastes, organic matrix waste forms, and cement waste forms containing a substantial component of organic waste. Irradiations were carried out by external gamma sources and by the incorporation of alpha emitters, such as 238 Pu. Irradiated materials included matrix materials, simulated waste forms and real waste forms. 2 figs.; 3 tabs.; 8 refs

  9. Results of field testing of waste forms using lysimeters

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  10. Disposal costs for SRP high-level wastes in borosilicate glass and crystalline ceramic waste forms

    International Nuclear Information System (INIS)

    Rozsa, R.B.; Campbell, J.H.

    1982-01-01

    Purpose of this document is to compare and contrast the overall burial costs of the glass and ceramic waste forms, including processing, storage, transportation, packaging, and emplacement in a repository. Amount of waste will require approximately 10,300 standard (24 in. i.d. x 9-5/6 ft length) canisters of waste glass, each containing about 3260 lb of waste at 28% waste loading. The ceramic waste form requires about one-third the above number of standard canisters. Approximately $2.5 billion is required to process and dispose of this waste, and the total cost is independent of waste form (glass or ceramic). The major cost items (about 80% of the total cost) for all cases are capital and operating expenses. The capital and 20-year operating costs for the processing facility are the same order of magnitude, and their sum ranges from about one-half of the total for the reference glass case to two-thirds of the total for the ceramic cases

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

    International Nuclear Information System (INIS)

    Mayberry, J.L.; Huebner, T.L.; Ross, W.; Nakaoka, R.; Schumacher, R.; Cunnane, J.; Singh, D.; Darnell, R.; Greenhalgh, W.

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-08-01

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

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

    International Nuclear Information System (INIS)

    Westsik, J.H. Jr.

    1984-08-01

    Incinerator ash from the combustion of general trash and ion exchange resins was immobilized in cement and bitumen. Tests were conducted on the resulting waste forms to provide a data base for the acceptability of actual low-level waste forms. The testing was done in accordance with the US Nuclear Regulatory Commission Technical Position on Waste Form. Bitumen had a measured compressive strength of 130 psi and a leachability index of 13 as measured with the ANS 16.1 leach test procedure. Cement demonstrated a compressive strength of 1400 psi and a leachability index of 7. Both waste forms easily exceed the minimum compressive strength of 50 psi and leachability index of 6 specified in the Technical Position. Irradiation to 10 8 Rad and exposure to 31 thermal cycles ranging from +60 0 ) to -30 0 C did not significantly impact these properties. Neither waste form supported bacterial or fungal growth as measured with ASTM G21 and G22 procedures. However, there is some indication of biodegradation due to co-metabolic processes. Concentration of organic complexants in leachates of the ash, cement and bitumen were too low to significantly affect the release of radionuclides from the waste forms. Neither bitumen nor cement containing incinerator ash caused any corrosion or degradation of potential container materials including steel, polyethylene and fiberglass. However, moist ash did cause corrosion of the steel

  14. Alternative solid forms for Savannah River Plant defense waste

    International Nuclear Information System (INIS)

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

    1980-01-01

    Solid forms and processes were evaluated for immobilization of SRP high-level radioactive waste, which contains bulk chemicals such as hydrous iron and aluminium oxides. Borosilicate glass currently is the best overall choice. High-silica glass, tailored ceramics, and coated ceramics are potentially superior products, but require more difficult processes

  15. Microstructural characterization of glass and ceramic simulated waste forms

    International Nuclear Information System (INIS)

    Headley, T.J.; Healey, J.T.; Hlava, P.F.; Kupfer, M.J.; Strachan, D.M.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  17. Zirconium phosphate waste forms for low-temperature stabilization of cesium-137-containing waste streams

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.S.; Tlustochowicz.

    1996-04-01

    Novel chemically bonded phosphate ceramics are being developed and fabricated for low-temperature stabilization and solidification of waste streams that are not amenable to conventional high-temperature stabilization processes because volatiles are present in the wastes. A composite of zirconium-magnesium phosphate has been developed and shown to stabilize ash waste contaminated with a radioactive surrogate of 137 Cs. Excellent retainment of cesium in the phosphate matrix system was observed in Toxicity Characteristic Leaching Procedure tests. This was attributed to the capture of cesium in the layered zirconium phosphate structure by intercalation ion-exchange reaction. But because zirconium phosphate has low strength, a novel zirconium/magnesium phosphate composite waste form system was developed. The performance of these final waste forms, as indicated by compression strength and durability in aqueous environments, satisfy the regulatory criteria. Test results indicate that zirconium-magnesium-phosphate-based final waste forms present a viable technology for treatment and solidification of cesium-contaminated wastes

  18. Testing of high-level waste forms under repository conditions

    International Nuclear Information System (INIS)

    Mc Menamin, T.

    1989-01-01

    The workshop on testing of high-level waste forms under repository conditions was held on 17 to 21 October 1988 in Cadarache, France, and sponsored by the Commission of the European Communities (CEC), the Commissariat a l'energie atomique (CEA) and the Savannah River Laboratory (US DOE). Participants included representatives from Australia, Belgium, Denmark, France, Germany, Italy, Japan, the Netherlands, Sweden, Switzerland, The United Kingdom and the United States. The first part of the conference featured a workshop on in situ testing of simulated nuclear waste forms and proposed package components, with an emphasis on the materials interface interactions tests (MIIT). MIIT is a sevent-part programme that involves field testing of 15 glass and waste form systems supplied by seven countries, along with potential canister and overpack materials as well as geologic samples, in the salt geology at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico, USA. This effort is still in progress and these proceedings document studies and findings obtained thus far. The second part of the meeting emphasized multinational experimental studies and results derived from repository systems simulation tests (RSST), which were performed in granite, clay and salt environments

  19. Radiation damage studies related to nuclear waste forms

    International Nuclear Information System (INIS)

    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 2 Ti 2 O 7 (pyrochlore) and CaZrTi 2 O 7 (zirconolite), of relative importance to current waste forms were studied independently by doping with 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 ΔV/V 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 2 Ti 2 O 7 and CaZrTi 2 O 7 . The later material exhibited anisotropic behavior in which the expansion of the monoclinic cell in the c 0 direction was over five times that of the a 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 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

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

    International Nuclear Information System (INIS)

    Mattus, C.H.

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mattus, C.H.

    2001-04-19

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

  2. 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.

  3. Salt-occluded zeolite waste forms: Crystal structures and transformability

    International Nuclear Information System (INIS)

    Richardson, J.W. Jr.

    1996-01-01

    Neutron diffraction studies of salt-occluded zeolite and zeolite/glass composite samples, simulating nuclear waste forms loaded with fission products, have revealed complex structures, with cations assuming the dual roles of charge compensation and occlusion (cluster formation). These clusters roughly fill the 6--8 angstrom diameter pores of the zeolites. Samples are prepared by equilibrating zeolite-A with complex molten Li, K, Cs, Sr, Ba, Y chloride salts, with compositions representative of anticipated waste systems. Samples prepared using zeolite 4A (which contains exclusively sodium cations) as starting material are observed to transform to sodalite, a denser aluminosilicate framework structure, while those prepared using zeolite 5A (sodium and calcium ions) more readily retain the zeolite-A structure. Because the sodalite framework pores are much smaller than those of zeolite-A, clusters are smaller and more rigorously confined, with a correspondingly lower capacity for waste containment. Details of the sodalite structures resulting from transformation of zeolite-A depend upon the precise composition of the original mixture. The enhanced resistance of salt-occluded zeolites prepared from zeolite 5A to sodalite transformation is thought to be related to differences in the complex chloride clusters present in these zeolite mixtures. Data relating processing conditions to resulting zeolite composition and structure can be used in the selection of processing parameters which lead to optimal waste forms

  4. 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.

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

    International Nuclear Information System (INIS)

    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

  6. Using mixture experiments to develop cementitious waste forms

    International Nuclear Information System (INIS)

    Spence, R.D.; Anderson, C.M.; Piepel, G.F.

    1993-01-01

    Mixture experiments are presented as a means to develop cementitious waste forms. The steps of a mixture experiment are (1) identifying the waste form ingredients; (2) determining the compositional constraints of these ingredients; (3) determining the extreme vertices, edge midpoints, and face centroids of the constrained multidimensional volume (these points along with some interior points represent the set of possible compositions for testing); (4) picking a subset of these points for the experimental design; (5) measuring the properties of the selected subset; and (6) generating the response surface models. The models provide a means for predicting the properties within the constrained region. This article presents an example of this process for one property: unconfined compressive strength

  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. Preliminary waste form characteristics report Version 1.0. Revision 1

    International Nuclear Information System (INIS)

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

    1991-01-01

    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

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

    International Nuclear Information System (INIS)

    Ken Bateman; Stephen Priebe

    2006-01-01

    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

  10. Support for DOE program in mineral waste-form development

    Energy Technology Data Exchange (ETDEWEB)

    Palmour, H. III; Hare, T.M.; Russ, J.C.; Batchelor, A.D.; Paisley, M.J.; Freed, L.E.

    1982-09-01

    This research investigation relates to sintered simulation ceramic waste forms of the generic SYNROC compositional type. Though they have been formulated with simulated wastes only, they serve as prototypes for potential hot, processed, crystalline waste forms whose combined thermodynamic stability and physical integrity are considered to render them capable of long-term imobilization of high-level radwastes under deep geologic disposal conditions. The problems involved are nontrivial, largely because of the very complex nature of the radwastes: a typical waste stream would contain more than 31 cation species. When the stabilizing matrix constituents are included, the final batch composition must successfully account (and find substitutional homes for some 35 different cation species. One of the important objectives of this study thus has been to develop a computer-based method for simulating these complex ion substitutions, and for calculating the resultant phase demands and batch formulations. Primary goals of the study have been (1) use of that computer simulation capability to incorporate rationally the radwaste ions from a specific waste stream (PW-7a) into the available SYNROC lattice sites and (2) utilization of existing ceramic processing and sintering methodologies to assure (and to understand) the attainment of high density, fine microstructure, full phase development and other features of the sintered product which are known to relate directly to its integrity and leach resistance. Though improved resistance to leaching has been a continuing goal, time and budget constraints have precluded initiation of any leachability studies of these new compositions during this contract period. 27 references, 15 figures, 6 tables.

  11. Support for DOE program in mineral waste-form development

    International Nuclear Information System (INIS)

    Palmour, H. III; Hare, T.M.; Russ, J.C.; Batchelor, A.D.; Paisley, M.J.; Freed, L.E.

    1982-09-01

    This research investigation relates to sintered simulation ceramic waste forms of the generic SYNROC compositional type. Though they have been formulated with simulated wastes only, they serve as prototypes for potential hot, processed, crystalline waste forms whose combined thermodynamic stability and physical integrity are considered to render them capable of long-term imobilization of high-level radwastes under deep geologic disposal conditions. The problems involved are nontrivial, largely because of the very complex nature of the radwastes: a typical waste stream would contain more than 31 cation species. When the stabilizing matrix constituents are included, the final batch composition must successfully account (and find substitutional homes for some 35 different cation species. One of the important objectives of this study thus has been to develop a computer-based method for simulating these complex ion substitutions, and for calculating the resultant phase demands and batch formulations. Primary goals of the study have been (1) use of that computer simulation capability to incorporate rationally the radwaste ions from a specific waste stream (PW-7a) into the available SYNROC lattice sites and (2) utilization of existing ceramic processing and sintering methodologies to assure (and to understand) the attainment of high density, fine microstructure, full phase development and other features of the sintered product which are known to relate directly to its integrity and leach resistance. Though improved resistance to leaching has been a continuing goal, time and budget constraints have precluded initiation of any leachability studies of these new compositions during this contract period. 27 references, 15 figures, 6 tables

  12. Thermal performance study of form-stable composite phase change material with polyacrylic

    Science.gov (United States)

    Kee, Shin Yiing; Munusamy, Yamuna; Ong, Kok Seng; Chee, Swee Yong; Sanmuggam, Shimalaa

    2017-04-01

    Phase change material (PCM) is one of the most popular and widely used as thermal energy storage material because it is able to absorb and release a large amount of latent heat during a phase change process over a narrow temperature range. In this work, the form-stable composite PCM was prepared by blending of PMMA and myristic acid in different weight percentage. PMMA was used as a supporting material while myristic acid was used as PCM. Theoretically, PCM can be encapsulated in the support material after blending. However, a small amount of liquid PCMs can leak out from supporting material due to the volume change in phase change process. Therefore, a form-stable composite PCM with polyacrylic coating was studied. Leakage test was carried out to determine the leakage percentage of the form-stable composite PCM. Fourier transform infrared spectroscopy (FTIR) was used to characterize the chemical compatibility of the form-stable PCM composite while differential scanning calorimetry (DSC) was used to study the melting, freezing point and the latent heat of melting and freezing for the form-stable composite PCM.

  13. Integrated Waste Management Strategy and Radioactive Waste Forms for the 21st Century

    International Nuclear Information System (INIS)

    Dirk Gombert; Jay Roach

    2007-01-01

    The U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) was announced in 2006. As currently envisioned, GNEP will be the basis for growth of nuclear energy worldwide, using a closed proliferation-resistant fuel cycle. The Integrated Waste Management Strategy (IWMS) is designed to ensure that all wastes generated by fuel fabrication and recycling will have a routine disposition path making the most of feedback to fuel and recycling operations to eliminate or minimize byproducts and wastes. If waste must be generated, processes will be designed with waste treatment in mind to reduce use of reagents that complicate stabilization and minimize volume. The IWMS will address three distinct levels of technology investigation and systems analyses and will provide a cogent path from (1) research and development (R and D) and engineering scale demonstration, (Level I); to (2) full scale domestic deployment (Level II); and finally to (3) establishing an integrated global nuclear energy infrastructure (Level III). The near-term focus of GNEP is on achieving a basis for large-scale commercial deployment (Level II), including the R and D and engineering scale activities in Level I that are necessary to support such an accomplishment. Throughout these levels is the need for innovative thinking to simplify, including regulations, separations and waste forms to minimize the burden of safe disposition of wastes on the fuel cycle

  14. Integrated Waste Management Strategy and Radioactive Waste Forms for the 21st Century

    Energy Technology Data Exchange (ETDEWEB)

    Dirk Gombert; Jay Roach

    2007-03-01

    The U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) was announced in 2006. As currently envisioned, GNEP will be the basis for growth of nuclear energy worldwide, using a closed proliferation-resistant fuel cycle. The Integrated Waste Management Strategy (IWMS) is designed to ensure that all wastes generated by fuel fabrication and recycling will have a routine disposition path making the most of feedback to fuel and recycling operations to eliminate or minimize byproducts and wastes. If waste must be generated, processes will be designed with waste treatment in mind to reduce use of reagents that complicate stabilization and minimize volume. The IWMS will address three distinct levels of technology investigation and systems analyses and will provide a cogent path from (1) research and development (R&D) and engineering scale demonstration, (Level I); to (2) full scale domestic deployment (Level II); and finally to (3) establishing an integrated global nuclear energy infrastructure (Level III). The near-term focus of GNEP is on achieving a basis for large-scale commercial deployment (Level II), including the R&D and engineering scale activities in Level I that are necessary to support such an accomplishment. Throughout these levels is the need for innovative thinking to simplify, including regulations, separations and waste forms to minimize the burden of safe disposition of wastes on the fuel cycle.

  15. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

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

    1980-07-01

    This volume contains 5 appendices. Title listing are: technologies for recovery of transuranics; nondestructive assay of TRU contaminated wastes; miscellaneous waste characteristics; acceptance criteria for TRU waste; and TRU waste treatment technologies

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

    International Nuclear Information System (INIS)

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

    1993-08-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

    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 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 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 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

  18. Preparation of plutonium waste forms with ICPP calcined high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Staples, B.A.; Knecht, D.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); O`Holleran, T.P. [Argonne National Lab.-West, Idaho Falls, ID (United States)] [and others

    1997-05-01

    Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce{sup +4}) as a surrogate for plutonium (Pu{sup +4}) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study.

  19. Preparation of plutonium waste forms with ICPP calcined high-level waste

    International Nuclear Information System (INIS)

    Staples, B.A.; Knecht, D.A.; O'Holleran, T.P.

    1997-05-01

    Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce +4 ) as a surrogate for plutonium (Pu +4 ) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  1. Characteristics of waste forms improved by using admixtures

    International Nuclear Information System (INIS)

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

    1989-06-01

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

  2. 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.

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

    International Nuclear Information System (INIS)

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

    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

  4. 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.

  5. Proposed waste form performance criteria and testing methods for low-level mixed waste

    International Nuclear Information System (INIS)

    Franz, E.M.; Fuhrmann, M.; Bowerman, B.; Bates, S.; Peters, R.

    1994-08-01

    This document describes proposed waste form performance criteria and testing method that could be used as guidance in judging viability of a waste form as a physico-chemical barrier to releases of radionuclides and RCRA regulated hazardous components. It is assumed that release of contaminants by leaching is the single most important property by which the effectiveness of a waste form is judged. A two-tier regimen is proposed. The first tier includes a leach test required by the Environmental Protection Agency and a leach test designed to determine the net forward leach rate for a variety of materials. The second tier of tests are to determine if a set of stresses (i.e., radiation, freeze-thaw, wet-dry cycling) on the waste form adversely impact its ability to retain contaminants and remain physically intact. It is recommended that the first tier tests be performed first to determine acceptability. Only on passing the given specifications for the leach tests should other tests be performed. In the absence of site-specific performance assessments (PA), two generic modeling exercises are described which were used to calculate proposed acceptable leach rates

  6. Production of sodalite waste forms by addition of glass

    International Nuclear Information System (INIS)

    Pereira, C.

    1995-01-01

    Spent nuclear fuel can be treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. Sodalite is one of the mineral waste forms under study. Fission products in the molten salt are ion-exchanged into zeolite A, which is converted to sodalite and consolidated. Sodalite can be formed directly from mixtures of salt and zeolite A at temperatures above 975 K; however, nepheline is usually produced as a secondary phase. Addition of small amounts of glass frit to the mixture reduced nepheline formation significantly. Loss of fission products was not observed for reaction below 1000 K. Hot-pressing of the sodalite powders yielded dense pellets (∼2.3 g/cm 3 ) without any loss of fission product species. Normalized release rates were below 1 g/m 2 ·day for pre-washed samples in 28-day leach tests based on standard MCC-1 tests but increased with the presence of free salt on the sodalite

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

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. 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.; Buchholtz ten Brink, Marilyn R.

    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.

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

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

    International Nuclear Information System (INIS)

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

    1996-12-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    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

  15. 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.

  16. 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.

  17. Stable emulsions formed by self-assembly of interfacial networks of dipeptide derivatives.

    Science.gov (United States)

    Bai, Shuo; Pappas, Charalampos; Debnath, Sisir; Frederix, Pim W J M; Leckie, Joy; Fleming, Scott; Ulijn, Rein V

    2014-07-22

    We demonstrate the use of dipeptide amphiphiles that, by hand shaking of a biphasic solvent system for a few seconds, form emulsions that remain stable for months through the formation of nanofibrous networks at the organic/aqueous interface. Unlike absorption of traditional surfactants, the interfacial networks form by self-assembly through π-stacking interactions and hydrogen bonding. Altering the dipeptide sequence has a dramatic effect on the properties of the emulsions formed, illustrating the possibility of tuning emulsion properties by chemical design. The systems provide superior long-term stability toward temperature and salts compared to with sodium dodecyl sulfate (SDS) and can be enzymatically disassembled causing on-demand demulsification under mild conditions. The interfacial networks facilitate highly tunable and stable encapsulation and compartmentalization with potential applications in cosmetics, therapeutics, and food industry.

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

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-04-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

    Borosilicate Glass Marbles and/or monoliths were rated among the top three waste forms for immobilization of all types of Hanford high-level waste. Supergrout Concrete and Bitumen, low temperature processes, are judged to be particularly suitable for immobilization and bulk disposal of high sodium blended wastes and/or residual liquid. This preliminary assessment indicates that certain ceramic waste forms (e.g., Tailored Ceramics, Supercalcine Ceramic, and SYNROC Ceramic) are equal to or superior to Borosilicate Glass waste forms for immobilization of Hanford sludges and radionuclides removed from salt cake and residual liquid. These ceramic waste forms can be made by the Sol Gel process. Some multibarrier waste forms (e.g., Coated Ceramics, Ceramic Pellets in Metal Matrix, and Glass in Metal Matrix) are judged to be superior waste forms for immobilization of Hanford sludges and/or radionuclide concentrate

  1. Preparation and properties of lauric acid/silicon dioxide composites as form-stable phase change materials for thermal energy storage

    Energy Technology Data Exchange (ETDEWEB)

    Fang Guiyin, E-mail: gyfang@nju.edu.cn [Department of Physics, Nanjing University, Hankou Road 22, Nanjing, Jiangsu 210093 (China); Li Hui [Department of Material Science and Engineering, Nanjing University, Nanjing 210093 (China); Liu Xu [Department of Physics, Nanjing University, Hankou Road 22, Nanjing, Jiangsu 210093 (China)

    2010-08-01

    Form-stable lauric acid (LA)/silicon dioxide (SiO{sub 2}) composite phase change materials were prepared using sol-gel methods. The LA was used as the phase change material for thermal energy storage, with the SiO{sub 2} acting as the supporting material. The structural analysis of these form-stable LA/SiO{sub 2} composite phase change materials was carried out using Fourier transformation infrared spectroscope (FT-IR). The microstructure of the form-stable composite phase change materials was observed by a scanning electronic microscope (SEM). The thermal properties and thermal stability were investigated by a differential scanning calorimeter (DSC) and a thermogravimetric analysis apparatus (TGA), respectively. The SEM results showed that the LA was well dispersed in the porous network of SiO{sub 2}. The DSC results indicated that the melting latent heat of the form-stable composite phase change material is 117.21 kJ kg{sup -1} when the mass percentage of the LA in the SiO{sub 2} is 64.8%. The results of the TGA showed that these materials have good thermal stability. The form-stable composite phase change materials can be used for thermal energy storage in waste heat recovery and solar heating systems.

  2. Preparation and properties of lauric acid/silicon dioxide composites as form-stable phase change materials for thermal energy storage

    International Nuclear Information System (INIS)

    Fang Guiyin; Li Hui; Liu Xu

    2010-01-01

    Form-stable lauric acid (LA)/silicon dioxide (SiO 2 ) composite phase change materials were prepared using sol-gel methods. The LA was used as the phase change material for thermal energy storage, with the SiO 2 acting as the supporting material. The structural analysis of these form-stable LA/SiO 2 composite phase change materials was carried out using Fourier transformation infrared spectroscope (FT-IR). The microstructure of the form-stable composite phase change materials was observed by a scanning electronic microscope (SEM). The thermal properties and thermal stability were investigated by a differential scanning calorimeter (DSC) and a thermogravimetric analysis apparatus (TGA), respectively. The SEM results showed that the LA was well dispersed in the porous network of SiO 2 . The DSC results indicated that the melting latent heat of the form-stable composite phase change material is 117.21 kJ kg -1 when the mass percentage of the LA in the SiO 2 is 64.8%. The results of the TGA showed that these materials have good thermal stability. The form-stable composite phase change materials can be used for thermal energy storage in waste heat recovery and solar heating systems.

  3. Reference waste form, basalts, and ground water systems for waste interaction studies

    Energy Technology Data Exchange (ETDEWEB)

    Deju, R.A.; Ledgerwood, R.K.; Long, P.E.

    1978-09-01

    This report summarizes the type of waste form, basalt, and ground water compositions to be used in theoretical and experimental models of the geochemical environment to be simulated in studying a typical basalt repository. Waste forms to be used in the experiments include, and are limited to, glass, supercalcine, and spent unreprocessed fuel. Reference basalts selected for study include the Pomona member and the Umtanum Unit, Shwana Member, of the Columbia River Basalt Group. In addition, a sample of the Basalt International Geochemical Standard (BCR-1) will be used for cross-comparison purposes. The representative water to be used is of a sodium bicarbonate composition as determined from results of analyses of deep ground waters underlying the Hanford Site. 12 figures, 13 tables.

  4. Reference waste form, basalts, and ground water systems for waste interaction studies

    International Nuclear Information System (INIS)

    Deju, R.A.; Ledgerwood, R.K.; Long, P.E.

    1978-09-01

    This report summarizes the type of waste form, basalt, and ground water compositions to be used in theoretical and experimental models of the geochemical environment to be simulated in studying a typical basalt repository. Waste forms to be used in the experiments include, and are limited to, glass, supercalcine, and spent unreprocessed fuel. Reference basalts selected for study include the Pomona member and the Umtanum Unit, Shwana Member, of the Columbia River Basalt Group. In addition, a sample of the Basalt International Geochemical Standard (BCR-1) will be used for cross-comparison purposes. The representative water to be used is of a sodium bicarbonate composition as determined from results of analyses of deep ground waters underlying the Hanford Site. 12 figures, 13 tables

  5. Silica based gel as a potential waste form for high level waste from fuel reprocessing

    International Nuclear Information System (INIS)

    Ford, C.E.; Dempster, T.J.; Melling, P.J.

    1983-10-01

    To assess the feasibility of safe disposal of high-level radioactive waste as synthetic clay, or material that would react with ground water to form clay, experiments have been carried out to determine the hydrothermal crystallisation and leaching behaviour of silica based gels fired at 900 deg C. Crystallisation rates at a pressure of 500 bars and at temperatures below 400 deg C are negligible and this more or less precludes pre-disposal production of synthetic clay on the scale required. Leaching experiments suggest that the leach rates of Cs from gels by distilled water are higher than those of boro-silicate glasses and SYNROC at the lower temperatures that would be preferred for geological storage. However, amounts of bulk dissolution of gels may be lower than those of boro-silicate glasses. The initial leaching behaviour of gels might be considerably improved by hot compaction at 900 to 1000 deg C. Consideration of likely waste form dissolution behaviour in a repository environment suggests that gels of appropriate composition might perform as well as, or better than, boro-silicate glasses. A novel hypothetical plant is described that could produce the gel waste form on the scale required on a more or less continuous basis. (author)

  6. Contaminant Release from Residual Waste in Closed Single-Shell Tanks and Other Waste Forms Associated with the Tanks

    International Nuclear Information System (INIS)

    Deutsch, William J.

    2008-01-01

    This chapter describes the release of contaminants from the various waste forms that are anticipated to be associated with closure of the single-shell tanks. These waste forms include residual sludge or saltcake that will remain in the tanks after waste retrieval. Other waste forms include engineered glass and cementitious materials as well as contaminated soil impacted by previous tank leaks. This chapter also describes laboratory testing to quantify contaminant release and how the release data are used in performance/risk assessments for the tank waste management units and the onsite waste disposal facilities. The chapter ends with a discussion of the surprises and lessons learned to date from the testing of waste materials and the development of contaminant release models

  7. 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.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  9. Advanced waste forms research and development. Annual report

    International Nuclear Information System (INIS)

    McCarthy, G.J.

    1976-01-01

    Thermogravimetric analysis on cesium aluminosilicates was performed. Pollucite, CsAlSi 2 O 6 , appears to have more than adequate thermal stability for Cs fixation in supercalcine. The possibility of leaving Ru as RuO 2 during supercalcine crystallization is being assessed. The apatite solid solution [A/sub ss/] phases used in supercalcine phase formation models have been easily prepared as pure phases. Use of Ca instead of Sr as an alkaline earth additive yields a more stable A/sub ss/ phase. Weight ratios have been optimized for two firing temperatures in terms of soxhlet leachability. However, this is not the preferred approach to supercalcine consolidation. The phase formation models developed for PW-4b-type wastes were routinely applied to PW-7, in spite of several significant composition differences in the latter. Uranium oxide remains as a fluorite structure oxide after supercalcine crystallization, but it does seem to have incorporated either or both CeO 2 or ZrO 2 . Characterization of this F/sub ss/ phase will require further study. Based on observations during several complex compatibility studies, it appears that good overall thermal stability at as high as 1200 0 C might be attainable for nepheline based supercalcine formulations. A thorough literature review of the oxide crystal chemistry of Te has been completed

  10. Preparation and thermal properties of form stable paraffin phase change material encapsulation

    International Nuclear Information System (INIS)

    Liu Xing; Liu Hongyan; Wang Shujun; Zhang Lu; Cheng Hua

    2006-01-01

    Paraffin waxes are cheap and have moderate thermal energy storage density but low thermal conductivity and, hence, require large surface area to be used in energy storage. Form stable paraffin phase change materials (PCM) in which paraffin serves as a latent heat storage material and polyolefins act as a supporting material, because of paraffin leakage, are required to be improved. The form stable paraffin PCM in the present paper was encapsulated in an inorganic silica gel polymer successfully by in situ polymerization. The differential scanning calorimeter (DSC) was used to measure its thermal properties. At the same time, the Washburn equation, which measures the wetting properties of powder materials, was used to test the hydrophilic-lipophilic properties of the PCMs. The result indicated that the enthalpy of the microencapsulated PCMs was reduced little, while their hydrophilic properties were enhanced largely

  11. 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.

  12. Progress in forming bottom barriers under waste sites

    International Nuclear Information System (INIS)

    Carter, E.E.

    1997-01-01

    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

  13. Low-risk alternative waste forms for problematic high-level and long-lived nuclear wastes

    International Nuclear Information System (INIS)

    Stewart, M.W.A.; Begg, B.D.; Moricca, S.; Day, R.A.

    2006-01-01

    Full text: The highest cost component the nuclear waste clean up challenge centres on high-level waste (HLW) and consequently the greatest opportunity for cost and schedule savings lies with optimising the approach to HLW cleanup. The waste form is the key component of the immobilisation process. To achieve maximum cost savings and optimum performance the selection of the waste form should be driven by the characteristics of the specific nuclear waste to be immobilised, rather than adopting a single baseline approach. This is particularly true for problematic nuclear wastes that are often not amenable to a single baseline approach. The use of tailored, high-performance, alternative waste forms that include ceramics and glass-ceramics, coupled with mature process technologies offer significant performance improvements and efficiency savings for a nuclear waste cleanup program. It is the waste form that determines how well the waste is locked up (chemical durability), and the number of repository disposal canisters required (waste loading efficiency). The use of alternative waste forms for problematic wastes also lowers the overall risk by providing high performance HLW treatment alternatives. The benefits tailored alternative waste forms bring to the HLW cleanup program will be briefly reviewed with reference to work carried out on the following: The HLW calcines at the Idaho National Laboratory; SYNROC ANSTO has developed a process utilising a glass-ceramic combined with mature hot-isostatic pressing (HIP) technology and has demonstrated this at a waste loading of 80 % and at a 30 kg HIP scale. The use of this technology has recently been estimated to result in a 70 % reduction in waste canisters, compared to the baseline borosilicate glass technology; Actinide-rich waste streams, particularly the work being done by SYNROC ANSTO with Nexia Solutions on the Plutonium-residues wastes at Sellafield in the UK, which if implemented is forecast to result in substantial

  14. Waste form performance assessment in the YUCCA Mountain engineered barrier system, American Nuclear Society

    International Nuclear Information System (INIS)

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

    2000-01-01

    This work demonstrates a technique for comparing the performance of waste forms in a repository environment when one or more of the waste forms constitute a small part of the total amount of waste planned for the repository. In applying the technique, it is important to identify radionuclides that are highly soluble in the transport fluid since it is only for these that the release is controlled by the dissolution rate of the waste form matrix. The techniques presented here have been applied to an evaluation of the performance of waste forms from the electrometallurgical treatment of spent fuel in the proposed Yucca Mountain Repository Engineered Barrier System (EBS)

  15. 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

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

    International Nuclear Information System (INIS)

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

    1980-09-01

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

  17. Radiation effects in glass waste forms for high-level waste and plutonium disposal

    International Nuclear Information System (INIS)

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

    1997-01-01

    A key challenge in the permanent disposal of high-level waste (HLW), plutonium residues/scraps, and excess weapons plutonium in glass waste forms is the development of predictive models of long-term performance that are based on a sound scientific understanding of relevant phenomena. Radiation effects from β-decay and α-decay can impact the performance of glasses for HLW and Pu disposition through the interactions of the α-particles, β-particles, recoil nuclei, and γ-rays with the atoms in the glass. Recently, a scientific panel convened under the auspices of the DOE Council on Materials Science to assess the current state of understanding, identify important scientific issues, and recommend directions for research in the area of radiation effects in glasses for HLW and Pu disposition. The overall finding of the panel was that there is a critical lack of systematic understanding on radiation effects in glasses at the atomic, microscopic, and macroscopic levels. The current state of understanding on radiation effects in glass waste forms and critical scientific issues are presented

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    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 Na 2 O that can be explained by more alkaline components nature – Na 2 SiO 3 , NaOH, and SO 3 . Fe 2 O 3 and ZnO are not detected in water extracts at all samples.

  19. Polymorphic drugs examined with neutron spectroscopy: Is making more stable forms really that simple?

    Energy Technology Data Exchange (ETDEWEB)

    Tsapatsaris, Nikolaos [European Spallation Source ESS AB, P.O. Box 176, SE-221 00 Lund (Sweden); Landsgesell, Sven [Helmholtz Zentrum Berlin, Hahn-Meitner Platz, 1, 14109 Berlin (Germany); Koza, Michael M.; Frick, Bernhard [Institut Laue-Langevin, BP 156 38042, Grenoble Cedex 9 (France); Boldyreva, Elena V. [Novosibirsk State University, ul. Pirogova 2, Novosibirsk 630090 (Russian Federation); Institute of Solid State Chemistry and Mechanochemistry, ul. Kutateladze 18, Novosibirsk 630128 (Russian Federation); Bordallo, Heloisa N., E-mail: bordallo@nbi.ku.dk [Niels Bohr Institute, University of Copenhagen, DK-2100 Copenhagen (Denmark)

    2013-12-12

    Highlights: • Pressure induced transition in the denser polymorphic form. • Weak strength of hydrogen bonds driving new stable packing. • Lattice anisotropy versus charge transfer. - Abstract: Understanding polymorphism in pharmaceutical ingredients is a long-standing challenge in formulation science. A well-known example is paracetamol, C{sub 8}H{sub 9}NO{sub 2}. The marketed stable form I crystallizes with corrugated molecular layers. In contrast, form II, which is thermodynamically favorable at high pressures, has relatively planar layers that can slip over each other without difficulty, but is metastable at ambient conditions. By means of inelastic neutron scattering we demonstrated that the lattice modes of form II exhibit a sudden 1 meV energy shift at 300 K under a pressure of ca 0.4 GPa. Moreover, evidence of an increase of the vibrational energy in both polymorphs was found, which was accompanied, in form I, by an unexpectedly weak increase of the tunnel splitting. These results indicate an anisotropy of the potential surface probed by the methyl rotor, and are discussed in relation to the differences of the strength of the hydrogen bond environment for each polymorph.

  20. 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.

  1. Immobilization of fission products in phosphate ceramic waste forms

    International Nuclear Information System (INIS)

    Singh, D.

    1996-01-01

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

  2. Activated carbon derived from waste coffee grounds for stable methane storage.

    Science.gov (United States)

    Kemp, K Christian; Baek, Seung Bin; Lee, Wang-Geun; Meyyappan, M; Kim, Kwang S

    2015-09-25

    An activated carbon material derived from waste coffee grounds is shown to be an effective and stable medium for methane storage. The sample activated at 900 °C displays a surface area of 1040.3 m(2) g(-1) and a micropore volume of 0.574 cm(3) g(-1) and exhibits a stable CH4 adsorption capacity of ∼4.2 mmol g(-1) at 3.0 MPa and a temperature range of 298 ± 10 K. The same material exhibits an impressive hydrogen storage capacity of 1.75 wt% as well at 77 K and 100 kPa. Here, we also propose a mechanism for the formation of activated carbon from spent coffee grounds. At low temperatures, the material has two distinct types with low and high surface areas; however, activation at elevated temperatures drives off the low surface area carbon, leaving behind the porous high surface area activated carbon.

  3. Activated carbon derived from waste coffee grounds for stable methane storage

    Science.gov (United States)

    Kemp, K. Christian; Baek, Seung Bin; Lee, Wang-Geun; Meyyappan, M.; Kim, Kwang S.

    2015-09-01

    An activated carbon material derived from waste coffee grounds is shown to be an effective and stable medium for methane storage. The sample activated at 900 °C displays a surface area of 1040.3 m2 g-1 and a micropore volume of 0.574 cm3 g-1 and exhibits a stable CH4 adsorption capacity of ˜4.2 mmol g-1 at 3.0 MPa and a temperature range of 298 ± 10 K. The same material exhibits an impressive hydrogen storage capacity of 1.75 wt% as well at 77 K and 100 kPa. Here, we also propose a mechanism for the formation of activated carbon from spent coffee grounds. At low temperatures, the material has two distinct types with low and high surface areas; however, activation at elevated temperatures drives off the low surface area carbon, leaving behind the porous high surface area activated carbon.

  4. Development of a ceramic waste form for high-level waste disposal

    International Nuclear Information System (INIS)

    Esh, D. W.

    1998-01-01

    A ceramic waste form is being developed by Argonne National Laboratory (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented

  5. Development of a ceramic waste form for high-level waste disposal

    International Nuclear Information System (INIS)

    Esh, D.W.; Goff, K.M.; Hirsche, K.T.; Battisti, T.J.; Simpson, M.F.; Johnson, S.G.; Bateman, K.J.

    1999-01-01

    A ceramic waste form is being developed by Argonne National Laboratory (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented. The process is demonstrated to be robust for the type and range of operation metrics considered and the performance metric (PCT durability test) against which the operation metrics are evaluated

  6. DNA CTG triplet repeats involved in dynamic mutations of neurologically related gene sequences form stable duplexes

    Science.gov (United States)

    Smith, G. K.; Jie, J.; Fox, G. E.; Gao, X.

    1995-01-01

    DNA triplet repeats, 5'-d(CTG)n and 5'-d(CAG)n, are present in genes which have been implicated in several neurodegenerative disorders. To investigate possible stable structures formed by these repeating sequences, we have examined d(CTG)n, d(CAG)n and d(CTG).d(CAG)n (n = 2 and 3) using NMR and UV optical spectroscopy. These studies reveal that single stranded (CTG)n (n > 2) forms stable, antiparallel helical duplexes, while the single stranded (CAG)n requires at least three repeating units to form a duplex. NMR and UV melting experiments show that the Tm increases in the order of [(CAG)3]2 DNA. However, unique NOE and 1H-31P coupling patterns associated with the repetitive T.T mismatches in the CTG repeats are discerned. These results, in conjunction with recent in vitro studies suggest that longer CTG repeats may form hairpin structures, which can potentially cause interruption in replication, leading to dynamic expansion or deletion of triplet repeats.

  7. 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. Copyright © 2014 Elsevier B.V. All rights reserved.

  8. Identification of items and activities important to waste form acceptance by Westinghouse GoCo sites

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Marra, S.L.; Dempster, J.; Randklev, E.H.

    1993-01-01

    The Department of Energy has established specifications (Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms, or WAPS) for canistered waste forms produced at Hanford, Savannah River, and West Valley. Compliance with these specifications requires that each waste form producer identify the items and activities which must be controlled to ensure compliance. As part of quality assurance oversight activities, reviewers have tried to compare the methodologies used by the waste form producers to identify items and activities important to waste form acceptance. Due to the lack of a documented comparison of the methods used by each producer, confusion has resulted over whether the methods being used are consistent. This confusion has been exacerbated by different systems of nomenclature used by each producer, and the different stages of development of each project. The waste form producers have met three times in the last two years, most recently on June 28, 1993, to exchange information on each producer's program. These meetings have been sponsored by the Westinghouse GoCo HLW Vitrification Committee. This document is the result of this most recent exchange. It fills the need for a documented comparison of the methodologies used to identify items and activities important to waste form acceptance. In this document, the methodology being used by each waste form producer is summarized, and the degree of consistency among the waste form producers is determined

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

    International Nuclear Information System (INIS)

    1981-01-01

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

  10. Waste form development and characterization in pyrometallurgical treatment of spent nuclear fuel

    International Nuclear Information System (INIS)

    Ackerman, J.

    1998-01-01

    Electrometallurgical treatment is a compact, inexpensive method that is being developed at Argonne National Laboratory to deal with spent nuclear fuel, primarily metallic and oxide fuels. In this method, metallic nuclear fuel constituents are electrorefined in a molten salt to separate uranium from the rest of the spent fuel. Oxide and other fuels are subjected to appropriate head end steps to convert them to metallic form prior to electrorefining. The treatment process generates two kinds of high-level waste--a metallic and a ceramic waste. Isolation of these wastes has been developed as an integral part of the process. The wastes arise directly from the electrorefiner, and waste streams do not contain large quantities of solvent or other process fluids. Consequently, waste volumes are small and waste isolation processes can be compact and rapid. This paper briefly summarizes waste isolation processes then describes development and characterization of the two waste forms in more detail

  11. G protein-coupled receptors form stable complexes with inwardly rectifying potassium channels and adenylyl cyclase.

    Science.gov (United States)

    Lavine, Natalie; Ethier, Nathalie; Oak, James N; Pei, Lin; Liu, Fang; Trieu, Phan; Rebois, R Victor; Bouvier, Michel; Hebert, Terence E; Van Tol, Hubert H M

    2002-11-29

    A large number of studies have demonstrated co-purification or co-immunoprecipitation of receptors with G proteins. We have begun to look for the presence of effector molecules in these receptor complexes. Co-expression of different channel and receptor permutations in COS-7 and HEK 293 cells in combination with co-immunoprecipitation experiments established that the dopamine D(2) and D(4), and beta(2)-adrenergic receptors (beta(2)-AR) form stable complexes with Kir3 channels. The D(4)/Kir3 and D(2) receptor/Kir3 interaction does not occur when the channel and receptor are expressed separately and mixed prior to immunoprecipitation, indicating that the interaction is not an artifact of the experimental protocol and reflects a biosynthetic event. The observed complexes are stable in that they are not disrupted by receptor activation or modulation of G protein alpha subunit function. However, using a peptide that binds Gbetagamma (betaARKct), we show that Gbetagamma is critical for dopamine receptor-Kir3 complex formation, but not for maintenance of the complex. We also provide evidence that Kir3 channels and another effector, adenylyl cyclase, are stably associated with the beta(2)-adrenergic receptor and can be co-immunoprecipitated by anti-receptor antibodies. Using bioluminescence resonance energy transfer, we have shown that in living cells under physiological conditions, beta(2)AR interacts directly with Kir3.1/3.4 and Kir3.1/3.2c heterotetramers as well as with adenylyl cyclase. All of these interactions are stable in the presence of receptor agonists, suggesting that these signaling complexes persist during signal transduction. In addition, we provide evidence that the receptor-effector complexes are also found in vivo. The observation that several G protein-coupled receptors form stable complexes with their effectors suggests that this arrangement might be a general feature of G protein-coupled signal transduction.

  12. Molybdenum and zinc stable isotope variation in mining waste rock drainage and waste rock at the Antamina mine, Peru

    Energy Technology Data Exchange (ETDEWEB)

    Skierszkan, E.K., E-mail: eskiersz@eos.ubc.ca [Department of Earth, Ocean and Atmospheric Sciences, University of British Columbia, 2020-2207 Main Mall, Vancouver V6T 1Z4 (Canada); Mayer, K.U. [Department of Earth, Ocean and Atmospheric Sciences, University of British Columbia, 2020-2207 Main Mall, Vancouver V6T 1Z4 (Canada); Weis, D. [Pacific Centre for Isotopic and Geochemical Research (PCIGR), Department of Earth, Ocean and Atmospheric Sciences, University of British Columbia, 2020-2207 Main Mall, Vancouver V6T 1Z4 (Canada); Beckie, R.D. [Department of Earth, Ocean and Atmospheric Sciences, University of British Columbia, 2020-2207 Main Mall, Vancouver V6T 1Z4 (Canada)

    2016-04-15

    The stable isotope composition of molybdenum (Mo) and zinc (Zn) in mine wastes at the Antamina Copper–Zn–Mo mine, Peru, was characterized to investigate whether isotopic variation of these elements indicated metal attenuation processes in mine drainage. Waste rock and ore minerals were analyzed to identify the isotopic composition of Mo and Zn sources, namely molybdenites (MoS{sub 2}) and sphalerites (ZnS). Molybdenum and Zn stable isotope ratios are reported relative to the NIST-SRM-3134 and PCIGR-1 Zn standards, respectively. δ{sup 98}Mo among molybdenites ranged from − 0.6 to + 0.6‰ (n = 9) while sphalerites showed no δ{sup 66}Zn variations (0.11 ± 0.01‰, 2 SD, n = 5). Mine drainage samples from field waste rock weathering experiments were also analyzed to examine the extent of isotopic variability in the dissolved phase. Variations spanned 2.2‰ in δ{sup 98}Mo (− 0.1 to + 2.1‰) and 0.7‰ in δ{sup 66}Zn (− 0.4 to + 0.3‰) in mine drainage over a wide pH range (pH 2.2–8.6). Lighter δ{sup 66}Zn signatures were observed in alkaline pH conditions, which was consistent with Zn adsorption and/or hydrozincite (Zn{sub 5}(OH){sub 6}(CO{sub 3}){sub 2}) formation. However, in acidic mine drainage Zn isotopic compositions reflected the value of sphalerites. In addition, molybdenum isotope compositions in mine drainage were shifted towards heavier values (0.89 ± 1.25‰, 2 SD, n = 16), with some overlap, in comparison to molybdenites and waste rock (0.13 ± 0.82‰, 2 SD, n = 9). The cause of heavy Mo isotopic signatures in mine drainage was more difficult to resolve due to isotopic heterogeneity among ore minerals and a variety of possible overlapping processes including dissolution, adsorption and secondary mineral precipitation. This study shows that variation in metal isotope ratios are promising indicators of metal attenuation. Future characterization of isotopic fractionation associated to key environmental reactions will improve the power

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

    International Nuclear Information System (INIS)

    Tait, J.C.

    1993-05-01

    AECL has investigated three waste forms for the immobilization of high-level liquid wastes that would arise if used CANDU fuels were reprocessed at some time in the future to remove fissile materials for the fabrication of new power reactor fuel. These waste forms are borosilicate glasses, aluminosilicate glasses and titanosilicate glass-ceramics. This report discusses the potential effects of alpha, beta and gamma radiation on the releases of radionuclides from these waste forms as a result of aqueous corrosion by groundwaters that would be present in an underground waste disposal vault. The report discusses solid-state damage caused by radiation-induced atomic displacements in the waste forms as well as irradiation of groundwater solutions (radiolysis), and their potential effects on waste-form corrosion and radionuclide release. The current literature on radiation effects on borosilicate glasses and in ceramics is briefly reviewed, as are potential radiation effects on specialized waste forms for the immobilization of 129 I, 85 Kr and 14 C. (author). 104 refs., 9 tabs., 5 figs

  14. 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

  15. 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

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

    International Nuclear Information System (INIS)

    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

  17. 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.

  18. 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.

  19. 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

  20. Modeling corrosion and constituent release from a metal waste form

    International Nuclear Information System (INIS)

    Bauer, T. H.; Fink, J. K.; Abraham, D. P.; Johnson, I.; Johnson, S. G.; Wigeland, R. A.

    2000-01-01

    Several ANL ongoing experimental programs have measured metal waste form (MWF) corrosion and constituent release. Analysis of this data has initiated development of a consistent and quantitative phenomenology of uniform aqueous MWF corrosion. The effort so far has produced a preliminary fission product and actinide release model based on measured corrosion rates and calibrated by immersion test data for a 90 C J-13 and concentrated J-13 solution environment over 1-2 year exposure times. Ongoing immersion tests of irradiated and unirradiated MWF samples using more aggressive test conditions and improved tracking of actinides will serve to further validate, modify, and expand the application base of the preliminary model-including effects of other corrosion mechanisms. Sample examination using both mechanical and spectrographic techniques will better define both the nature and durability of the protective barrier layer. It is particularly important to assess whether the observations made with J-13 solution at 900 C persist under more aggressive conditions. For example, all the multiplicative factors in Table 1 implicitly assume the presence of protective barriers. Under sufficiently aggressive test conditions, such protective barriers may very well be altered or even eliminated

  1. The development of mechanically formed stable nanobubbles intended for sonoporation-mediated gene transfection.

    Science.gov (United States)

    Abdalkader, Rodi; Kawakami, Shigeru; Unga, Johan; Higuchi, Yuriko; Suzuki, Ryo; Maruyama, Kazuo; Yamashita, Fumiyoshi; Hashida, Mitsuru

    2017-11-01

    In this study, stable nano-sized bubbles (nanobubbles [NBs]) were produced using the mechanical agitation method in the presence of perfluorocarbon gases. NBs made with perfluoropropane had a smaller size (around 400 nm) compared to that of those made with perfluorobutane or nitrogen gas. The lipid concentration in NBs affected both their initial size and post-formulation stability. NBs formed with a final lipid concentration of 0.5 mg/ml tended to be more stable, having a uniform size distribution for 24 h at room temperature and 50 h at 4 °C. In vitro gene expression revealed that NBs/pDNA in combination with ultrasound (US) irradiation had significantly higher transfection efficacy in colon C26 cells. Moreover, for in vivo gene transfection in mice left limb muscles, there was notable local transfection activity by NBs/pDNA when combined with US irradiation. In addition, the aged NBs kept at room temperature or 4 °C were still functional at enhancing gene transfection in mice. We succeeded in preparing stable NBs for efficient in vivo gene transfection, using the mechanical agitation method.

  2. 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...

  3. 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.

  4. 40 CFR 761.345 - Form of the waste to be sampled.

    Science.gov (United States)

    2010-07-01

    ..., AND USE PROHIBITIONS Sampling Non-Liquid, Non-Metal PCB Bulk Product Waste for Purposes of Characterization for PCB Disposal in Accordance With § 761.62, and Sampling PCB Remediation Waste Destined for Off-Site Disposal, in Accordance With § 761.61 § 761.345 Form of the waste to be sampled. PCB bulk product...

  5. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

    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

  6. Immobilization of fission products in phosphate ceramic waste forms

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.

    1997-01-01

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

  7. Quality checking task force destructive testing of active waste forms

    International Nuclear Information System (INIS)

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

    1987-03-01

    The implications of sampling and testing of full size active packages of intermediate level wastes are summarised in this report. Sampling operations are technically feasible but a major difficulty will be the disposal of secondary waste. A literature survey indicated that destructive testing of wasteforms is not carried out as a routine operation in Europe or the USA. (author)

  8. 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.

  9. Indefinitely stable iron(IV) cage complexes formed in water by air oxidation

    Science.gov (United States)

    Tomyn, Stefania; Shylin, Sergii I.; Bykov, Dmytro; Ksenofontov, Vadim; Gumienna-Kontecka, Elzbieta; Bon, Volodymyr; Fritsky, Igor O.

    2017-01-01

    In nature, iron, the fourth most abundant element of the Earth's crust, occurs in its stable forms either as the native metal or in its compounds in the +2 or +3 (low-valent) oxidation states. High-valent iron (+4, +5, +6) compounds are not formed spontaneously at ambient conditions, and the ones obtained synthetically appear to be unstable in polar organic solvents, especially aqueous solutions, and this is what limits their studies and use. Here we describe unprecedented iron(IV) hexahydrazide clathrochelate complexes that are assembled in alkaline aqueous media from iron(III) salts, oxalodihydrazide and formaldehyde in the course of a metal-templated reaction accompanied by air oxidation. The complexes can exist indefinitely at ambient conditions without any sign of decomposition in water, nonaqueous solutions and in the solid state. We anticipate that our findings may open a way to aqueous solution and polynuclear high-valent iron chemistry that remains underexplored and presents an important challenge.

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

    International Nuclear Information System (INIS)

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

    1982-08-01

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

  11. DEVELOPMENT OF MODEL FOR QUANTITATIVE EVALUATION OF DYNAMICALLY STABLE FORMS OF RIVER CHANNELS

    Directory of Open Access Journals (Sweden)

    O. V. Zenkin

    2017-01-01

    Full Text Available The article highlights the method of calculating the optimum curvature of the river channels using the kinematic model of the flow structure based on the concept of discrete nature of the channel process. It offers the analytic form of the equation of motion of river flow, which can be used simulation modeling for searching dynamically stable form of the river channel, and which can control water level in rivers. The source data for the illustrations of given in the article modeling methods have been served the images received from MODIS on the Terra satellite, for the lower reaches of the river Kur, which merges with the river Urmi, forming the Tunguska river – the left tributary of the Amur.The modified geometric method can be used to calculate obliquity of tangent to the curve and normal in those situations when observed on satellite imagery points are located on the coordinate of the network irregularly and when three points lying on the curve of the riverbed do not form isosceles triangle.The model assembles tangential and radial components of the forces acting on the water flow (centrifugal, friction and gravity. Curvature radius is explicitly expressed in the model through the parameter  – gradient angle relative to the axis X. As solution for the value of the angle  is searched, when the correlation function reaches its maximum. It is assumed that the riverbed shape “wrong” and could be modified so that the resulting curve better correlated with calculated curve. Morphometric dependences for macroforms allow creating series of morphological methods for the calculation of deformations and displacement of the shore in any section of meander scroll.The proposed technique has been tested also on satellite imagery of high resolution. The presented methods of calculation are used as the basis for hydrological projects of geoinformation systems oriented at prediction of morphodynamic processes and morphological evolution of river

  12. Molecular Environmental Science Using Synchrotron Radiation: Chemistry and Physics of Waste Form Materials. Final Report

    International Nuclear Information System (INIS)

    Lindle, Dennis W.

    2011-01-01

    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.

  13. Evaluation of remote smearing of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Payne, C.H.; Rankin, W.N.

    1991-01-01

    The Savannah River Site (SRS) is evaluating the variables of the remote smearing process for monitoring transferable contamination on the waste glass canisters at the Defense Waste Processing Facility (DWPF). Smearing for transferable contamination is typically done by hand, but in this case, due to the nature of the high level waste within the canisters, remote smearing is required. The effectiveness of the smear pad was determined under varying conditions (distance traveled, force applied, and canister surface), as well as the relative importance of these factors. It was concluded that the remote smear is more reliable than the hand smear

  14. An experimental survey of the factors that affect leaching from low-level radioactive waste forms

    International Nuclear Information System (INIS)

    Dougherty, D.R.; Pietrzak, R.F.; Fuhrmann, M.; Colombo, P.

    1988-09-01

    This report represents the results of an experimental survey of the factors that affect leaching from several types of solidified low-level radioactive waste forms. The goal of these investigations was to determine those factors that accelerate leaching without changing its mechanism(s). Typically, although not in every case,the accelerating factors include: increased temperature, increased waste loading (i.e., increased waste to binder ratio), and decreased size (i.e., decreased waste form volume to surface area ratio). Additional factors that were studied were: increased leachant volume to waste form surface area ratio, pH, leachant composition (groundwaters, natural and synthetic chelating agents), leachant flow rate or replacement frequency and waste form porosity and surface condition. Other potential factors, including the radiation environment and pressure, were omitted based on a survey of the literature. 82 refs., 236 figs., 13 tabs

  15. An experimental survey of the factors that affect leaching from low-level radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Dougherty, D.R.; Pietrzak, R.F.; Fuhrmann, M.; Colombo, P.

    1988-09-01

    This report represents the results of an experimental survey of the factors that affect leaching from several types of solidified low-level radioactive waste forms. The goal of these investigations was to determine those factors that accelerate leaching without changing its mechanism(s). Typically, although not in every case,the accelerating factors include: increased temperature, increased waste loading (i.e., increased waste to binder ratio), and decreased size (i.e., decreased waste form volume to surface area ratio). Additional factors that were studied were: increased leachant volume to waste form surface area ratio, pH, leachant composition (groundwaters, natural and synthetic chelating agents), leachant flow rate or replacement frequency and waste form porosity and surface condition. Other potential factors, including the radiation environment and pressure, were omitted based on a survey of the literature. 82 refs., 236 figs., 13 tabs.

  16. A minichaperone-based fusion system for producing insoluble proteins in soluble stable forms.

    Science.gov (United States)

    Sharapova, Olga A; Yurkova, Maria S; Fedorov, Alexey N

    2016-02-01

    We have developed a fusion system for reliable production of insoluble hydrophobic proteins in soluble stable forms. A carrier is thermophilic minichaperone, GroEL apical domain (GrAD), a 15 kDa monomer able to bind diverse protein substrates. The Met-less variant of GrAD has been made for further convenient use of Met-specific CNBr chemical cleavage, if desired. The Met-less GrAD retained stability and solubility of the original protein. Target polypeptides can be fused to either C-terminus or N-terminus of GrAD. The system has been tested with two unrelated insoluble proteins fused to the C-terminus of GrAD. One of the proteins was also fused to GrAD N-terminus. The fusions formed inclusion bodies at 25°C and above and were partly soluble only at lower expression temperatures. Most importantly, however, after denaturation in urea, all fusions without exception were completely renatured in soluble stable forms that safely survived freezing-thawing as well as lyophilization. All fusions for both tested target proteins retained solubility at high concentrations for days. Functional analysis revealed that a target protein may retain functionality in the fusion. Convenience features include potential thermostability of GrAD fusions, capacity for chemical and enzymatic cleavage of a target and His6 tag for purification. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please e-mail: journals.permissions@oup.com.

  17. A stable "flat" form of two-dimensional crystals: could graphene, silicene, germanene be minigap semiconductors?

    Science.gov (United States)

    O'Hare, A; Kusmartsev, F V; Kugel, K I

    2012-02-08

    The discovery of a flat two-dimensional crystal known as graphene has contradicted Landau-Peierls-Mermin-Wagner arguments that there is no stable flat form of such crystals. Here, we show that the "flat" shape of graphene arises due to a microscopic buckling at the smallest possible interatomic scale. We show that the graphene, silicene, and other two-dimensional crystals are stable due to transverse short-range displacements of appropriate atoms. The distortions are small and form various patterns, which we describe in a framework of Ising model with competing interactions. We show that when temperature decreases, two transitions, disorder into order and order into disorder, arise. The ordered state has a form of stripes where carbon atoms are shifted regularly with respect to the plane. The flat graphene, silicene, or germanene planes look like a microscopic "washboard" with the wavelength of about couple of interatomic spacing of appropriate sublattices, which for graphene is about 1.8-3.6 Å. At lower temperatures, the ordered state transforms into a glass. Because of up-down asymmetry in buckled graphene, silicene and other two-dimensional crystals deposited on substrate, a minibandgap may arise. We derive a criterion for the minigap formation and show how it is related to the buckling and to the graphene-substrate interaction. Because of the bandgap, there may arise new phenomena and in particular a rectification of ac current induced by microwave or infrared radiation. We show that the amplitude of direct current arising at wave mixing of two harmonics of microwave electromagnetic radiation is huge. Moreover, we predict the existence of miniexcitons and a new type of fermionic minipolaritons whose behavior can be controlled by the microwave and terahertz radiation. © 2012 American Chemical Society

  18. Characterization and durability testing of a glass-bonded ceramic waste form

    International Nuclear Information System (INIS)

    Johnson, S. G.

    1998-01-01

    Argonne National Laboratory is developing a glass bonded ceramic waste form for encapsulating the fission products and transuranics from the conditioning of metallic reactor fuel. This waste form is currently being scaled to the multi-kilogram size for encapsulation of actual high level waste. This paper will present characterization and durability testing of the ceramic waste form. An emphasis on results from application of glass durability tests such as the Product Consistency Test and characterization methods such as X-ray diffraction and scanning electron microscopy. The information presented is based on a suite of tests utilized for assessing product quality during scale-up and parametric testing

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

    International Nuclear Information System (INIS)

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

    1994-03-01

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

  20. 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).

  1. Assessment of spent-fuel waste-form/stabilizer alternatives for geologic disposal

    International Nuclear Information System (INIS)

    Einziger, R.E.; Himes, D.A.

    1982-06-01

    The Office of Nuclear Waste Isolation (ONWI) is studying the possibility of burying canisterized unreprocessed spent fuel in a deep geologic repository. One aspect of this study is an assessment of the possible spent fuel waste forms. The fuel performance portion of the Waste Form Assessment was to evaluate five candidate spent fuel waste forms for postemplacement performance with emphasis on their ability to retard the release of radionuclides to the repository geology. Spent fuel waste forms under general consideration were: (1) unaltered fuel assembly; (2) fuel assembly with end fittings removed to shorten the length; (3 rods vented to remove gases and resealed; (4) disassembled fuel bundles to close-pack the rods; and (5) rods chopped and fragments immobilized in a matrix material. Thirteen spent fuel waste forms, classified by generic stabilizer type, were analyzed for relative in-repository performance based on: (1) waste form/stabilizer support against lithostatic pressure; (2) long-term stability for radionuclide retention; (3) minimization of cladding degradation; (4) prevention of canister/repository breach due to pressurization; (5) stabilizer heat transfer; (6) the stabilizer as an independent barrier to radionuclide migration; and (7) prevention of criticality. The waste form candidates were ranked as follows: (1) the best waste form/stabilizer combination is the intact assembly, with or without end bells, vented (and resealed) or unvented, with a solid stabilizer; (2) a suitable alternative is the combination of bundled close-packed rods with a solid stabilizer around the outside of the bundle to resist lithostatic pressure; and (3) the other possible waste forms are of lower ranking with the worst waste form/stabilizer combination being the intact assembly with a gas stabilizer or the chopped fuel

  2. Waste acceptance product specifications for vitrified high-level waste forms

    International Nuclear Information System (INIS)

    Applewhite-Ramsey, A.; Sproull, J.F.

    1994-01-01

    The Department of Energy (DOE) Office of Environmental Restoration and Waste Management (EM) has developed Waste Acceptance Product Specifications (EM-WAPS). The EM-WAPS will be the basis for defining product acceptance criteria compatible with the requirements of the Civilian Radioactive Waste Management System (CRWMS). The relationship between the EM-WAPS and the CRWMS Systems Requirements document (WA-SRD) will be discussed. The impact of the EM-WAPS on the Savannah River Sit (SRS) Defense Waste Processing Facility's (DWPF) Waste Acceptance Program, Waste Qualification Run planning, and startup schedule will also be reported. 14 refs., 2 tabs

  3. PWR Users Group 10 CFR 61 Waste Form Requirements Compliance Test Program

    International Nuclear Information System (INIS)

    Rosenlof, R.C.

    1985-01-01

    In January of 1984, a PWR Users Group was formed to initiate a 10 CFR 61 Waste Form Requirements Compliance Test Program on a shared cost basis. The original Radwaste Solidification Systems sold by ATCOR ENGINEERED SYSTEMS, INC. to the utilities were required to produce a free-standing monolith with no free water. None of the other requirements of 10 CFR 61 had to be met. Current regulations, however, have substantially expanded the scope of the waste form acceptance criteria. These new criteria required that generators of radioactive waste demonstrate the ability to produce waste forms which meet certain chemical and physical requirements. This paper will present the test program used and the results obtained to insure 10 CFR 61 compliance of the three (3) typical waste streams generated by the ATCOR PWR Users Group's plants. The primary objective of the PWR Users Group was not to maximize waste loading within the masonry cement solidification media, but to insure that the users Radwaste Solidification System is capable of producing waste forms which meet the waste form criteria of 10 CFR 61. A description of the laboratory small sample certification program and the actual full scale pilot plant verification approach used is included in this paper. Also included is a discussion of the development of a Process Control Program to ensure the reproducibility of the test results with actual waste

  4. 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).

  5. 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).

  6. Correlation between microstructure and mechanical properties of stable mixtures formed by austenite and martensite

    International Nuclear Information System (INIS)

    Eckstein, C.B.

    1982-03-01

    The influence of martensite in mechanical properties of stable mixtures formed by austenite and martensite was studied by varying the amount of martensite in the mixtures. Microstructural parameters were determined by Optical Quantitative Metallography and used to establish the correlation between the mechanical response of the mixtures in tension and their microstructures. The 'in situ' deformation of each phase in mixtures was determined experimentally in terms of the rule of mixtures. It is shown that the partitioning of the deformation depends on the amount of martensite in the mixture and that it tends to a condition of isostrain at higher martensite volume fractions. Optical observation of fractured specimens showed that the beginning of the fracture process may related to regions of the austenite grain boundaries where they meet martensite plates. (Author) [pt

  7. Study on polyethylene glycol/epoxy resin composite as a form-stable phase change material

    International Nuclear Information System (INIS)

    Fang Yutang; Kang Huiying; Wang Weilong; Liu Hong; Gao Xuenong

    2010-01-01

    Form-stable polyethylene glycol (PEG)/epoxy resin (EP) composite as a novel phase change material (PCM) was prepared using casting molding method. In this new material, PEG acts as the latent heat storage material and EP polymer serves as the supporting material, which provides structural strength and prevents the leakage of the melted PEG. The structure and morphology of the novel composite were observed using Fourier transformation infrared spectroscope (FTIR) and scanning electronic microscope (SEM). The thermo-mechanical property and transition behavior were characterized by polarizing optical microscope (POM), static thermo-mechanical analysis (TMA) and differential scanning calorimeter (DSC). The experimental results show that, as a result of the physical tangled function of the epoxy resin carrier to the PEG segment, the composite macroscopically presents the solid-solid phase change characteristic.

  8. Leaching characteristics of the metal waste form from the electrometallurgical treatment process: Product consistency testing

    International Nuclear Information System (INIS)

    Johnson, S. G.; Keiser, D. D.; Frank, S. M.; DiSanto, T.; Noy, M.

    1999-01-01

    Argonne National Laboratory is developing an electrometallurgical treatment for spent fuel from the experimental breeder reactor II. A product of this treatment process is a metal waste form that incorporates the stainless steel cladding hulls, zirconium from the fuel and the fission products that are noble to the process, i.e., Tc, Ru, Nb, Pd, Rh, Ag. The nominal composition of this waste form is stainless steel/15 wt% zirconium/1--4 wt% noble metal fission products/1--2 wt % U. Leaching results are presented from several tests and sample types: (1) 2 week monolithic immersion tests on actual metal waste forms produced from irradiated cladding hulls, (2) long term (>2 years) pulsed flow tests on samples containing technetium and uranium and (3) crushed sample immersion tests on cold simulated metal waste form samples. The test results will be compared and their relevance for waste form product consistency testing discussed

  9. The effect of actinides on the microstructural development in a metallic high-level nuclear waste form

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Sinkler, W.; Abraham, D.P.; Richardson, J.W. Jr.; McDeavitt, S.M.

    1999-01-01

    Waste forms to contain material residual from an electrometallurgical treatment of spent nuclear fuel have been developed by Argonne National Laboratory. One of these waste forms contains waste stainless steel (SS), fission products that are noble to the process (e.g., Tc, Ru, Pd, Rh), Zr, and actinides. The baseline composition of this metallic waste form is SS-15wt.% Zr. The metallurgy of this baseline alloy has been well characterized. On the other hand, the effects of actinides on the alloy microstructure are not well understood. As a result, SS-Zr alloys with added U, Pu, and/or Np have been cast and then characterized, using scanning electron microscopy, transmission electron microscopy, and neutron diffraction, to investigate the microstructural development in SS-Zr alloys that contain actinides. Actinides were found to congregate non-uniformally in a Zr(Fe,Cr,Ni) 2+x phase. Apparently, the actinides were contained in varying amounts in the different polytypes (C14, C15, and C36) of the Zr(Fe,Cr,Ni) 2+x phase. Heat treatment of an actinide-containing SS-15 wt.% Zr alloy showed the observed microstructure to be stable

  10. 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.

  11. FY-87 packing fabrication techniques (commercial waste form) results

    International Nuclear Information System (INIS)

    Werry, E.V.; Gates, T.E.; Cabbage, K.S.; Eklund, J.D.

    1988-04-01

    This report covers the investigation of fabrication techniques associated with the development of suitable materials and methods to provide a prefabricated packing for waste packages for the Basalt Waste Isolation Project (BWIP). The principal functions of the packing are to minimize container corrosion during the 300 to 1000 years following repository closure and provide long-term control of the release of radionuclides from the waste package. The investigative work, discussed in this report, was specifically conceived to develop the design criteria for production of full-scale prototypical packing rings. The investigative work included the preparation of procedures, the preparation of fabrication materials, physical properties, and the determination of the engineering properties. The principal activities were the preparation of the materials and the determination of the physical properties. 21 refs., 20 figs., 14 tabs

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  13. Advanced waste forms research and development. Comprehensive progress report

    International Nuclear Information System (INIS)

    McCarthy, G.J.

    1977-01-01

    Progress is reported in the following research areas: cesium fixation in aluminosilicates and supercalcine development; low sodium wastes and high sodium wastes. Phase relations in the appropriate (for supercalcine) portion of the Cs 2 O-Al 2 O 3 -SiO 2 system are being investigated. Several compatibility lines (tie lines) have been established on the phase diagram. Methods of preparing pure single phases of CsAlSiO 4 and CsAlSi 2 O 6 (pollucite) are being developed. Compatibility studies continue to be the basis of supercalcine crystalline phase formation modeling

  14. Hanford Site Composite Analysis Technical Approach Description: Waste Form Release.

    Energy Technology Data Exchange (ETDEWEB)

    Hardie, S. [CH2M HILL Plateau Remediation Company, Richland, WA (United States); Paris, B. [CH2M HILL Plateau Remediation Company, Richland, WA (United States); Apted, M. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

    2017-09-14

    The U.S. Department of Energy (DOE) in DOE O 435.1 Chg. 1, Radioactive Waste Management, requires the preparation and maintenance of a composite analysis (CA). The primary purpose of the CA is to provide a reasonable expectation that the primary public dose limit is not likely to be exceeded by multiple source terms that may significantly interact with plumes originating at a low-level waste disposal facility. The CA is used to facilitate planning and land use decisions that help assure disposal facility authorization will not result in long-term compliance problems; or, to determine management alternatives, corrective actions or assessment needs, if potential problems are identified.

  15. Modeling the degradation of a metallic waste form intended for geologic disposal

    International Nuclear Information System (INIS)

    Bauer, T.H.; Morris, E.E.

    2007-01-01

    Nuclear reactors operating with metallic fuels have led to development of robust metallic waste forms intended to immobilize hazardous constituents in oxidizing environments. Release data from a wide range of tests where small waste form samples have been immersed in a variety of oxidizing solutions have been analyzed and fit to a mechanistically-derived 'logarithmic growth' form for waste form degradation. A bounding model is described which plausibly extrapolates these fits to long-term degradation in a geologic repository. The resulting empirically-fit degradation model includes dependence on solution pH, temperature, and chloride concentration as well as plausible estimates of statistical uncertainty. (authors)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

    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

  17. Sampling and transport of paraffin waste form from CWDS of nuclear power plant

    International Nuclear Information System (INIS)

    Lee, J. M.; Hwang, J. H.; Kim, C. R.; Park, J. W.

    2000-01-01

    Sampling and transport of paraffin waste form from concentrated waste drying system (CWDS) of domestic nuclear power plant were performed to collect the leaching characteristic data for the disposal of radioactive waste. Transport was performed according to the national regulations and the internal rules of the nuclear power plant. The sample of paraffin waste form was classified as L type package according to the regulation and radiation exposure of operator was measured in the range of 6 to 12 mrem that was less than the estimated amount

  18. 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

  19. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  3. Development, evaluation, and selection of candidate high-level waste forms

    International Nuclear Information System (INIS)

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

    1982-01-01

    The seven candidate waste forms, evaluated as potential media for the immobilization and gelogic disposal of high-level nuclear wastes were borosilicate glass, SYNROC, tailored ceramic, high-silica glass, FUETAP concrete, coated sol-gel particles, and glass marbles in a lead matrix. The evaluation, completed on August 1, 1981, combined preliminary waste form evaluations conducted at Department of Energy (DOE) defense waste-sites and at 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, respectively, for continued development and evaluation in the National HLW Program. The borosilicate glass and ceramic forms were further compared during FY-1982 on the basis of risk assessments, cost comparisons, properties comparisons, and conformance with proposed regulatory and repository criteria. Both the glass and ceramic forms are viable candidates for use at DOE defense HLW sites; they are also candidates for immobilization of commercial reprocessing wastes. This paper describes the waste form screening process, discusses each of the four major inputs considered in the selection of the two forms in 1981, and presents a brief summary of the comparisons of the two forms during 1982 and the selection process to determine the final form for SRP defense HLW

  4. 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.

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

    International Nuclear Information System (INIS)

    Serne, R. Jeffrey; Westsik, Joseph H.

    2011-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Abotsi, G.M.K. [Clark Atlanta Univ., GA (United States); Bostick, D.T.; Beck, D.E. [Oak Ridge National Lab., TN (United States)] [and others

    1996-05-01

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

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

    International Nuclear Information System (INIS)

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

    1996-05-01

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

  8. Experimental comparison of alternative solid forms for Savannah River high-level wastes

    International Nuclear Information System (INIS)

    Stone, J.A.

    1981-01-01

    Some of the conclusions of this study are: no waste form has a clear advantage; ceramic forms retain uranium best; high-silica glass retains cesium best; all forms retain rare earths (cerium) well; all forms leach incongruently; leaching increases with temperature; effect of leachant is small; and effect of sludge type is small, with exceptions

  9. NNWSI waste form testing at Argonne National Laboratory; Semiannual report: January-June 1987

    Energy Technology Data Exchange (ETDEWEB)

    Bates, J.K.; Gerding, T.J.; Abrajano, T.A. Jr.; Ebert, W.L.; Mazer, J.J. [Argonne National Lab., IL (USA)

    1988-11-01

    The Nevada Nuclear Waste Storage Investigation (NNWSI) Project is investigating the tuff beds of Yucca Mountain, Nevada, as a potential location for a high-level radioactive waste repository. As part of the waste package development portion of this project, experiments are being performed by the Chemical Technology Division of Argonne National Laboratory to study the behavior of the waste form under anticipated repository conditions. These experiments include the development and performance of a test to measure waste form behavior in unsaturated conditions and the performance of experiments designed to study the behavior of waste package components in an irradiated environment. Previous reports document developments in these areas through 1986. This report summarizes progress during the period January--June 1987, 19 refs., 17 figs., 20 tabs.

  10. State-of-the-art review of materials properties of nuclear waste forms

    International Nuclear Information System (INIS)

    Mendel, J.E.; Nelson, R.D.; Turcotte, R.P.; Gray, W.J.; Merz, M.D.; Roberts, F.P.; Weber, W.J.; Westsik, J.H. Jr.; Clark, D.E.

    1981-04-01

    The Materials Characterization Center (MCC) was established at the Pacific Northwest Laboratory to assemble a standardized nuclear waste materials data base for use in research, systems and facility design, safety analyses, and waste management decisions. This centralized data base will be provided through the means of a Nuclear Waste Materials Handbook. The first issue of the Handbook will be published in the fall of 1981 in looseleaf format so that it can be updated as additional information becomes available. To ensure utmost reliability, all materials data appearing in the Handbook will be obtained by standard procedures defined in the Handbook and approved by an independent Materials Review Board (MRB) comprised of materials experts from Department of Energy laboratories and from universities and industry. In the interim before publication of the Handbook there is need for a report summarizing the existing materials data on nuclear waste forms. This review summarizes materials property data for the nuclear waste forms that are being developed for immobilization of high-level radioactive waste. It is intended to be a good representation of the knowledge concerning the properties of HLW forms as of March 1981. The table of contents lists the following topics: introduction which covers waste-form categories, and important waste-form materials properties; physical properties; mechanical properties; chemical durability; vaporization; radiation effects; and thermal phase stability

  11. Process and equipment qualification of the ceramic and metal waste forms for spent fuel treatment

    International Nuclear Information System (INIS)

    Marsden, Ken; Knight, Collin; Bateman, Kenneth; Westphal, Brian; Lind, Paul

    2005-01-01

    The electrometallurgical process for treating sodium-bonded spent metallic fuel at the Materials and Fuels Complex of the Idaho National Laboratory separates actinides and partitions fission products into two waste forms. The first is the metal waste form, which is primarily composed of stainless steel from the fuel cladding. This stainless steel is alloyed with 15w% zirconium to produce a very corrosion-resistant metal which binds noble metal fission products and residual actinides. The second is the ceramic waste form which stabilizes fission product-loaded chloride salts in a sodalite and glass composite. These two waste forms will be packaged together for disposal at the Yucca Mountain repository. Two production-scale metal waste furnaces have been constructed. The first is in a large argon-atmosphere glovebox and has been used for equipment qualification, process development, and process qualification - the demonstration of process reliability for production of the DOE-qualified metal waste form. The second furnace will be transferred into a hot cell for production of metal waste. Prototype production-scale ceramic waste equipment has been constructed or procured; some equipment has been qualified with fission product-loaded salt in the hot cell. Qualification of the remaining equipment with surrogate materials is underway. (author)

  12. 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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  14. Performance of high level waste forms and engineered barriers under repository conditions

    International Nuclear Information System (INIS)

    1991-02-01

    The IAEA initiated in 1977 a co-ordinated research programme on the ''Evaluation of Solidified High-Level Waste Forms'' which was terminated in 1983. As there was a continuing need for international collaboration in research on solidified high-level waste form and spent fuel, the IAEA initiated a new programme in 1984. The new programme, besides including spent fuel and SYNROC, also placed greater emphasis on the effect of the engineered barriers of future repositories on the properties of the waste form. These engineered barriers included containers, overpacks, buffer and backfill materials etc. as components of the ''near-field'' of the repository. The Co-ordinated Research Programme on the Performance of High-Level Waste Forms and Engineered Barriers Under Repository Conditions had the objectives of promoting the exchange of information on the experience gained by different Member States in experimental performance data and technical model evaluation of solidified high level waste forms, components of the waste package and the complete waste management system under conditions relevant to final repository disposal. The programme includes studies on both irradiated spent fuel and glass and ceramic forms as the final solidified waste forms. The following topics were discussed: Leaching of vitrified high-level wastes, modelling of glass behaviour in clay, salt and granite repositories, environmental impacts of radionuclide release, synroc use for high--level waste solidification, leachate-rock interactions, spent fuel disposal in deep geologic repositories and radionuclide release mechanisms from various fuel types, radiolysis and selective leaching correlated with matrix alteration. Refs, figs and tabs

  15. Process description and plant design for preparing ceramic high-level waste forms

    International Nuclear Information System (INIS)

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

    1983-01-01

    The ceramics process flow diagram has been simplified and upgraded to utilize only two major processing steps - fluid-bed calcination and hot isostatic press consolidating. Full-scale fluid-bed calcination has been used at INEL to calcine high-level waste for 18 y; and a second-generation calciner, a fully remotely operated and maintained calciner that meets ALARA guidelines, started calcining high-level waste in 1982. Full-scale hot isostatic consolidation has been used by DOE and commercial enterprises to consolidate radioactive components and to encapsulate spent fuel elements for several years. With further development aimed at process integration and parametric optimization, the operating knowledge of full-scale demonstration of the key process steps should be rapidly adaptable to scale-up of the ceramic process to full plant size. Process flowsheets used to prepare ceramic and glass waste forms from defense and commercial high-level liquid waste are described. Preliminary layouts of process flow diagrams in a high-level processing canyon were prepared and used to estimate the preliminary cost of the plant to fabricate both waste forms. The estimated costs for using both options were compared for total waste management costs of SRP high-level liquid waste. Using our design, for both the ceramic and glass plant, capital and operating costs are essentially the same for both defense and commercial wastes, but total waste management costs are calculated to be significantly less for defense wastes using the ceramic option. It is concluded from this and other studies that the ceramic form may offer important advantages over glass in leach resistance, waste loading, density, and process flexibility. Preliminary economic calculations indicate that ceramics must be considered a leading candidate for the form to immobilize high-level wastes

  16. Alternatives for high-level waste forms, containers, and container processing systems

    International Nuclear Information System (INIS)

    Crawford, T.W.

    1995-01-01

    This study evaluates alternatives for high-level waste forms, containers, container processing systems, and onsite interim storage. Glass waste forms considered are cullet, marbles, gems, and monolithic glass. Small and large containers configured with several combinations of overpack confinement and shield casks are evaluated for these waste forms. Onsite interim storage concepts including canister storage building, bore holes, and storage pad were configured with various glass forms and canister alternatives. All favorable options include the monolithic glass production process as the waste form. Of the favorable options the unshielded 4- and 7-canister overpack options have the greatest technical assurance associated with their design concepts due to their process packaging and storage methods. These canisters are 0.68 m and 0.54 m in diameter respectively and 4.57 m tall. Life-cycle costs are not a discriminating factor in most cases, varying typically less than 15 percent

  17. Plutonium and surrogate fission products in a composite ceramic waste form

    International Nuclear Information System (INIS)

    Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T.

    1999-01-01

    Argonne National Laboratory is developing a ceramic waste form to immobilize salt containing fission products and transuranic elements. Preliminary results have been presented for ceramic waste forms containing surrogate fission products such as cesium and the lanthanides. In this work results from scanning electron microscopy/energy dispersive spectroscopy and x-ray diffraction are presented in greater detail for ceramic waste forms containing surrogate fission products. Additionally, results for waste forms containing plutonium and surrogate fission products are presented. Most of the surrogate fission products appear to be silicates or aluminosilicates whereas the plutonium is usually found in an oxide form. There is also evidence for the presence of plutonium within the sodalite phase although the chemical speciation of the plutonium is not known

  18. Form And Aging Of Plutonium In Savannah River Site Waste Tank 18

    International Nuclear Information System (INIS)

    Hobbs, D.

    2012-01-01

    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) 4(am) or hydrous plutonium oxide, PuO 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 4+ that has been substituted for other metal ions in crystal lattice sites, Pu 4+ occluded within hydrous metal oxide particles and Pu 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 2(am,hyd) is also likely to be

  19. Evaluation of a radioactive concrete waste form recovered from an ocean dumpsite

    International Nuclear Information System (INIS)

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

    1982-01-01

    Little dissolution of the concrete waste form in the ocean environment occurred as evidenced by a maximum waste package weight loss of approximately 5%. Water loss through evaporation during curing and dissolution of calcium hydroxide in disposal or inaccuracy of the initial weighing are believed to be responsible for the apparent weight loss. A conservative estimate that assumes a constant 0.33%/yr weight loss due solely to cement-phase dissolution predicts that it would require a minimum of 300 years in this environment before the concrete waste form would lose its integrity. The measured compression strength of the concrete waste form is in the range expected for concrete formulations. This indicates the absence of appreciable attack which is also supported by the observation that negligible deterioration of the waste form surface has occurred. The concrete waste form contained Cs-137, Cs-134, and Co-60. Based on the assumed initial Cs-137 distribution in the waste form, a bulk leach rate for this radionuclide of 2.4x10 -3 g/(cm 2 -day) was calculated. This corresponds to an average fractional activity loss rate of 3.7x10 -2 per year (neglecting decay). 7 figures, 1 table

  20. 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.

  1. Sensitivity tests of the waste-form-alone design for the low-activity-waste disposal system

    International Nuclear Information System (INIS)

    Fayer, M.J.; White, M.D.; Kincaid, C.T.

    1997-09-01

    Computer simulations were performed to assess the performance of the waste-form-alone (WFA) design for the low-activity-waste (LAW) disposal system. In FY 1997, PNNL performed additional simulations for Lockheed Martin Hanford Company (LMHC) to address specific questions about the disposal. LMHC manages the Glass Performance Assessment Project for DOE. The objectives of the additional simulations were to demonstrate the impact of grid resolution, diffusion, fracture flow within the waste form, and consumption of water by the waste form. It was assumed that the waste form would be the only part of the engineered disposal system that inhibits radionuclide release, referred to as WFA assumption. All calculations were performed with the latest version of the STOMP (Subsurface Transport Over Multiple Phases) simulator. Multiple simulations of the WFA disposal were performed to identify parameter and conceptual model sensitivities. The corrosion rate, recharge rate, well interception factor, hydraulic properties, and hydraulic and retardation models were shown to be important. Diffusion was shown to be important for the gravel model of glass but not the soil model. The impact of temperature changes was discussed and determined to be negligible. Water consumption during corrosion was evaluated and found to have a minimal effect on the dose calculations. Fracture flow within the glass was evaluated and found to have only a minimal effect on the dose calculations

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pet' kov, Vladimir; Asabina, Elena, E-mail: e.a.asabina@gmail.com; Loshkarev, Vladimir; Sukhanov, Maksim

    2016-04-01

    We have summarized our data and literature ones on the thermophysical properties and hydrolytic stability of Sr{sub 0.5}Zr{sub 2}(PO{sub 4}){sub 3} compound as a host NaZr{sub 2}(PO{sub 4}){sub 3}-type (NZP) structure for immobilization of {sup 90}Sr-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 Sr{sub 0.5}Zr{sub 2}(PO{sub 4}){sub 3}. Calculated values of thermal conductivity coefficients at zero porosity in the range 298–673 K were 1.86–2.40 W·m{sup −1} K{sup −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 Sr{sup 2+} 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 Sr{sup 2+} ions into distilled water. Feeble sinterability constrains practical applications of NZP substances, that is why known in literature methods of Sr{sub 0.5}Zr{sub 2}(PO{sub 4}){sub 3} dense ceramics obtaining have been reviewed. - Graphical abstract: The ability of Sr{sub 0.5}Zr{sub 2}(PO{sub 4}){sub 3} using as a perspective nuclear waste form, resistant to radiation damage, has been investigated. The obtained results highlight several concerns with application of Sr{sub 0.5}Zr{sub 2}(PO{sub 4}){sub 3} material as a host matrix, which is stable to extreme environmental condition action and able to include 90Sr radioisotope into insoluble framework structure. - Highlights: • The ability of Sr{sub 0.5}Zr{sub 2}(PO{sub 4}){sub 3} using as a nuclear waste form was investigated. • Its heat capacity, thermal expansion and thermal conductivity were studied. • Its stability in distilled water, aggressive (acid and alkaline) media was

  4. Development and testing of matrices for the encapsulation of glass and ceramic nuclear waste forms

    International Nuclear Information System (INIS)

    Wald, J.W.; Brite, D.W.; Gurwell, W.E.; Buckwalter, C.Q.; Bunnell, L.R.; Gray, W.J.; Blair, H.T.; Rusin, J.M.

    1982-02-01

    This report details the results of research on the matrix encapsulation of high level wastes at PML over the past few years. The demonstrations and tests described were designed to illustrate how the waste materials are effected when encapsulated in an inert matrix. Candidate materials evaluated for potential use as matrices for encapslation of pelletized ceramics or glass marbles were categorized into four groups: metals, glasses, ceramics, and graphite. Two processing techniques, casting and hot pressing, were investigated as the most promising methods of formation or densification of the matrices. The major results reported deal with the development aspects. However, chemical durability tests (leach tests) of the matrix materials themselves and matrix-waste form composites are also reported. Matrix waste forms can provide a low porosity, waste-free barrier resulting in increased leach protection, higher impact strength and improved thermal conductivity compared to unencapsulated glass or ceramic waste materials. Glass marbles encapsulated in a lead matrix offer the most significant improvement in waste form stability of all combinations evaluated. This form represents a readily demonstrable process that provides high thermal conductivity, mechanical shock resistance, radiation shielding and increased chemical durability through both a chemical passivation mechanism and as a physical barrier. Other durable matrix waste forms evaluated, applicable primarily to ceramic pellets, involved hot-pressed titanium or TiO 2 materials. In the processing of these forms, near 100% dense matrices were obtained. The matrix materials had excellent compatibility with the waste materials and superior potential chemical durability. Cracking of the hot-pressed ceramic matrix forms, in general, prevented the realization of their optimum properties

  5. 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.

  6. Testing and evaluation of solidified high-level waste forms

    International Nuclear Information System (INIS)

    Engelmann, C.

    1984-01-01

    The report describes research by several laboratories on the behaviour, in aqueous and salt environments, of borosilicate glass ceramics proposed for the solidification of nuclear wastes by the European Community. Results were obtained on inactive simulates, doped materials, and on borosilicate glass containing real radioactive waste. The influence of many important parameters were studied: leaching mode, nature of the leachant, pH, pressure, temperature, duration of the treatment, etc. The results of tests lasting for as little as a few hours or for as long as several hundred days, at temperatures up to 200 0 C or under pressures up to 200 bars, are presented. Numerous analytical techniques (ESCA, EMP, IRR, SEM, etc.) were used to determine the structure and the chemical composition of the altered layer developed by hydration at the glass surface. Information is also given on physical properties of the borosilicate glass: crystallization phase separation, alpha-irradiation stability, mechanical and thermal stability, etc. Finally, preliminary results on the structure and composition of hollandite ceramics are given

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

    International Nuclear Information System (INIS)

    Burnay, S.G.

    1987-05-01

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

  8. 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

  9. Three-dimensional thermal analysis of in-floor type nuclear waste repository for a ceramic waste form

    International Nuclear Information System (INIS)

    Sizgek, G. Devlet

    2005-01-01

    A thermal model is constructed and analyses are performed for an 'in-floor' type nuclear waste repository in granitic rock for a high level nuclear waste (HLW)-bearing ceramic waste form (synroc). Transient calculations for a three-dimensional (3-D) model have been carried out for both 20 and 10 wt.% HLW-bearing synroc, for surface cooling periods between reactor discharge and geological disposal varying from 5 to 40 years. This study investigates the temperature distribution in one of the boreholes of a hypothetical tunnel for a basic geometrical setting as well as the effect of varying the distance between adjacent boreholes and the distance between adjacent tunnels. The temperatures in the repository were found to be sensitive to the interim surface cooling period as well as the amount of waste loaded. The results showed that decreasing the spacing between the canisters has a more pronounced effect on the temperature field than decreasing the spacing between the tunnels

  10. Stabilities of nuclear waste forms and their geochemical interactions in repositories

    International Nuclear Information System (INIS)

    White, W.B.

    1980-01-01

    The stabilities of high-level nuclear waste forms in a repository environment are briefly discussed. The advantages and disadvantages of such waste forms as borosilicate glass, supercalcine ceramics, and synthetic minerals are presented in context with the different rock types which have been proposed as possible host rocks for repositories. It is concluded that the growing geochemical evidence favors the use of a silicate rock repository because of the effectiveness of aluminosilicate rocks as chemical barriers for most radionuclides

  11. Development of stable solidification methods for toxic lead oxide in radioactive wastes

    International Nuclear Information System (INIS)

    Hitoshi Mimura; Shingo Ikeda; Yuichi Niibori

    2009-01-01

    The objective of this study is to develop the advanced solidification methods for the toxic lead oxide contained in radioactive wastes and to examine their chemical durability in terms of leachability and surface alteration; the solidification characteristics and leachability for the following three kinds of solidified products immobilizing lead were examined, and the experimental results were summarized as follows. (a) Mineral solidified products: A-zeolite or fly ash (FA) was used as a binder, and NaAlO 2 and Na 2 SiO 3 were mixed as additives. The leachability of lead ions in pure water was considerably lowered by the heat treatment at higher temperature (1,000 degree C), and the concentration of lead ions leached was under criterion value of 0.3 mg/l. The products prepared by mixing A-zeolite and fly ash also had low leachability under 0.3 mg/l even in the saturated Ca(OH)2 solution. (b) Melted solidified products: A-zeolite or fly ash was used as a binder and glass-forming reagents of B 2 O 3 and NaH 2 PO 4 were used as additives. The XRD peaks assigned PbO were not observed in all products. The products for the mixtures of FA:NaH 2 PO 4 :PbO (2:2:1 and 3:1:1) had low leachability under criterion value in both leachants of deionized water and saturated Ca(OH) 2 solution. (c) Phosphate ceramics products: the chemically bonded phosphate ceramics were produced by using MgKPO 4 , MgHPO 4 , Zr(HPO 4 ) 2 , potassium iron phosphate and sodium iron phosphate, and FA was used as additives. In particular, by using MgHPO 4 , the leachability of the products was lowered less than 0.3 mg/l in both leachants. The phosphate ceramics products and melted solidified products are favorable as the waste solid forms immobilizing lead. In particular, novel ceramics products have advantages in the simple solidification procedure similarly to the cement products. As for mineral solidification, natural zeolites and FA as binder also useful from the viewpoint of cost efficiency

  12. 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.

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

    International Nuclear Information System (INIS)

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

    1982-03-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. Capric-myristic acid/vermiculite composite as form-stable phase change material for thermal energy storage

    Energy Technology Data Exchange (ETDEWEB)

    Karaipekli, Ali; Sari, Ahmet [Department of Chemistry, Gaziosmanpasa University, 60240 Tokat (Turkey)

    2009-03-15

    Phase change materials (PCMs) can be incorporated with building materials to obtain novel form-stable composite PCM which has effective energy storage performance in latent heat thermal energy storage (LHTES) systems. In this study, capric acid (CA)-myristic acid (MA) eutectic mixture/vermiculite (VMT) composite was prepared as a novel form-stable PCM using vacuum impregnation method. The composite PCM was characterized using scanning electron microscope (SEM) and Fourier transformation infrared (FT-IR) analysis technique. Thermal properties and thermal reliability of the composite PCM were determined by differential scanning calorimetry (DSC) analysis. The CA-MA eutectic mixture could be retained by 20 wt% into pores of the VMT without melted PCM seepage from the composite and therefore, this mixture was described as form-stable composite PCM. Thermal cycling test showed that the form-stable composite PCM has good thermal reliability and chemical stability although it was subjected to 3000 melting/freezing cycling. Thermal conductivity of the form-stable CA-MA/VMT composite PCM was increased by about 85% by introducing 2 wt% expanded graphite (EG) into the composite. The increase in thermal conductivity was confirmed by comparison of the melting and freezing times of the CA-MA/VMT composite with that of CA-MA/VMT/EG composite. The form-stable PCM including EG can be used as energy absorbing building material such as lightweight aggregate for plaster, concrete compounds, fire stop mortar, and component of interior fill for wallboards or hollow bricks because of its good thermal properties, thermal and chemical reliability and thermal conductivity. (author)

  16. 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

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

    International Nuclear Information System (INIS)

    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

  18. The effects of aging on compressive strength of low-level radioactive waste form samples

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Neilson, R.M. Jr.

    1996-06-01

    The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program, funded by the US Nuclear Regulatory Commission (NRC), is (a) studying the degradation effects in organic ion-exchange resins caused by radiation, (b) examining the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of 10 CFR 61 using solidified ion-exchange resins, (c) obtaining performance information on solidified ion-exchange resins in a disposal environment, and (d) determining the condition of liners used to dispose ion-exchange resins. Compressive tests were performed periodically over a 12-year period as part of the Technical Position testing. Results of that compressive testing are presented and discussed. During the study, both portland type I-II cement and Dow vinyl ester-styrene waste form samples were tested. This testing was designed to examine the effects of aging caused by self-irradiation on the compressive strength of the waste forms. Also presented is a brief summary of the results of waste form characterization, which has been conducted in 1986, using tests recommended in the Technical Position on Waste Form. The aging test results are compared to the results of those earlier tests. 14 refs., 52 figs., 5 tabs

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    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.

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

    International Nuclear Information System (INIS)

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

    1993-10-01

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

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

    International Nuclear Information System (INIS)

    Indacochea, J. E.; Gattu, V. K.; Chen, X.; Rahman, T.

    2017-01-01

    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

  2. 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

  3. Investigation of processing advanced waste forms in Cold Crucible Induction Melter systems

    Energy Technology Data Exchange (ETDEWEB)

    Roach, Jay A. [Nexergy Technical, LLC, Falls Church, Virginia (United States)

    2013-07-01

    The Cold Crucible Induction Melter (CCIM) technology has been identified in several independent reviews and studies as a potential alternative or supplemental vitrification technology for specific U.S. Department of Energy (DOE) high level radioactive waste (HLW) streams. These inventories have unique chemistries, such as high alumina, high iron, high chromium, etc., that are not amenable to efficient immobilization in conventional borosilicate glass (BSG) matrices. Advanced waste forms, such as iron phosphate glass (FePG), high crystalline BSG (HC-BSG), and alumino-silicate glass-ceramic (GC) have been shown to provide significantly improved immobilization matrices for these challenging waste streams, including dramatically improved waste loadings; however, they are difficult to efficiently process in the baseline joule-heated ceramic melters (JHCMs). The CCIM technology can effectively process these advanced waste forms at the desired high waste loadings. This paper describes the most recent collaborative activities between the DOE Office of Environmental Management (DOE EM) and various Russian institutions in modeling, development and testing of advanced waste forms in CCIMs, including FePG and HC-BSG formulations. Additionally, past interactions with other international partners, including France and Korea, are described. Finally, recent enhancements to CCIM processing related to modeling, draining techniques and automated control, which were collaboratively developed, will be discussed. (authors)

  4. 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

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

    International Nuclear Information System (INIS)

    Bleier, A.

    1997-09-01

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

  6. Preparation and flammability of high density polyethylene/paraffin/organophilic montmorillonite hybrids as a form stable phase change material

    International Nuclear Information System (INIS)

    Cai, Yibing; Hu, Yuan; Song, Lei; Kong, Qinghong; Yang, Rui; Zhang, Yinping; Chen, Zuyao; Fan, Weicheng

    2007-01-01

    A kind of form stable phase change material (PCM) based on high density polyethylene (HDPE), paraffin, organophilic montmorillonite (OMT) and intumescent flame retardant (IFR) hybrids is prepared by using a twin screw extruder technique. This kind of form stable PCM is made of paraffin as a dispersed phase change material and HDPE as a supporting material. The structure of the montmorillonite (MMT) and OMT is characterized by X-ray diffraction (XRD) and high resolution electron microscopy (HREM). The analysis indicates that the MMT is a kind of lamellar structure, and the structure does not change after organic modification. However, the structure of the hybrid is evidenced by the XRD and scanning electronic microscope (SEM). Its thermal stability, latent heat and flame retardant properties are given by the Thermogravimetry analysis (TGA), differential scanning calorimeter (DSC) method and cone calorimeter, respectively. Synergy is observed between the OMT and IFR. The XRD result indicates that the paraffin intercalates into the silicate layers of the OMT, thus forming a typically intercalated hybrid. The SEM investigation and DSC result show that the additives of OMT and IFR have hardly any effect on the HDPE/paraffin three dimensional netted structure and the latent heat. In TGA curves, although the onset of weight loss of flame-retardant form stable PCMs occur at a lower temperature than that of form stable PCM, flame-retardant form stable PCMs produce a large amount of char residue at 700 o C. The synergy between OMT and IFR leads to the decrease of the heat release rate (HRR), contributing to improvement of the flammability performance

  7. Evaluation of low and intermediate level radioactive solidified waste forms and packages

    International Nuclear Information System (INIS)

    1990-10-01

    Evaluation of low and intermediate level radioactive waste forms and packages with respect to compliance with quality and safety requirements for transport, interim storage and disposal has become a very important part of the radioactive waste management strategy in many countries. The evaluation of waste forms and packages provides precise basic data for regulatory bodies to establish safety requirements, and implement quality control and quality assurance procedures for radioactive waste management programmes. The requirements depend very much upon the disposal option selected, treatment technology used, waste form characteristics, package quality and other factors. The regulatory requirements can also influence the methodology of waste form/package evaluation together with selection and analysis of data for quality control and safety assurance. A coordinated research programme started at the end of 1985 and brought together 12 participants from 11 countries. The results of the programme and each particular project were discussed at three Research Coordination Meetings held in Cairo, Egypt, in May, 1986; in Beijing, China, in April, 1998; and at Harwell Laboratory, United Kingdom, in November, 1989. This document summarises the salient features and results achieved during the four year investigation and a recommendation for future work in this area. Refs, figs and tabs

  8. Epsilon metal waste form for immobilization of noble metals from used nuclear fuel

    Science.gov (United States)

    Crum, Jarrod V.; Strachan, Denis; Rohatgi, Aashish; Zumhoff, Mac

    2013-10-01

    Epsilon metal (ɛ-metal), an alloy of Mo, Pd, Rh, Ru, and Tc, is being developed as a waste form to treat and immobilize the undissolved solids and dissolved noble metals from aqueous reprocessing of commercial used nuclear fuel. Epsilon metal is an attractive waste form for several reasons: increased durability relative to borosilicate glass, it can be fabricated without additives (100% waste loading), and in addition it also benefits borosilicate glass waste loading by eliminating noble metals from the glass, thus the processing problems related to their insolubility in glass. This work focused on the processing aspects of the epsilon metal waste form development. Epsilon metal is comprised of refractory metals resulting in high alloying temperatures, expected to be 1500-2000 °C, making it a non-trivial phase to fabricate by traditional methods. Three commercially available advanced technologies were identified: spark-plasma sintering, microwave sintering, and hot isostatic pressing, and investigated as potential methods to fabricate this waste form. Results of these investigations are reported and compared in terms of bulk density, phase assemblage (X-ray diffraction and elemental analysis), and microstructure (scanning electron microscopy).

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-12-30

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

  11. 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.

  12. An alternative waste form for the final disposal of high-level radioactive waste (HLW) on the basis of a survey of solidification and final disposal of HLW

    International Nuclear Information System (INIS)

    Bauer, C.

    1982-01-01

    The dissertation comprises two separate parts. The first part presents the basic conditions and concepts of the process leading to the development of a waste form, such as:origin, composition and characteristics of the high-level radioactive waste; evaluation of the methods available for the final disposal of radioactive waste, especially the disposal in a geological formation, including the resulting consequences for the conditions of state in the surroundings of the waste package; essential option for the conception of a waste form and presentation of the waste forms developed and examined on an international level up to now. The second part describes the production of a waste form on TiO 2 basis, in which calcined radioactive waste particles in the submillimeter range are embedded in a rutile matrix. That waste form is produced by uniaxial pressure sintering in the temperature range of 1223 K to 1423 K and pressures between 5 MPa and 20 MPa. Microstructure, mechanical properties and leaching rates of the waste form are presented. Moreover, a method is explained allowing compacting of the rutile matrix and also integration of a wasteless overpack of titanium or TiO 2 into the waste form. (orig.) [de

  13. 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).

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

    International Nuclear Information System (INIS)

    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)

  15. From non-disposable to disposable, treatment of pyrophoric or gas forming waste forms for disposal - Thermal treatment of pyrophoric or gas-forming metals

    International Nuclear Information System (INIS)

    Oesterberg, Carl; Lindberg, Maria

    2014-01-01

    In order to dispose of waste in either a deep geological disposal or in a shallower repository there are several demands that the waste and its package must fulfil, one is that it is not to react with oxygen or the waste package or backfill in the repository, i.e. concrete or grout. The waste forms that do not fulfil this particular criterion must be treated in some way to render the waste non-reactive. One of these waste are metallic uranium. Metallic uranium is not only an issue originating from the nuclear industry, as old types of fuel, it is also present in, for example, transport flasks and as samples used in schools, which all has to be disposed of sooner or later. Another waste that arise is magnesium doped with thorium, originating from the aviation, aerospace and missile industry. These alloys are now being replaced with others without thorium so they are in need of handling and possibly treatment before disposal. Magnesium metal is also pyrophoric, in particular in molten or powder form. In order to evaluate thermally treating these metals in a very controlled environment, such as a pyrolysis vessel, experimental work has been performed. The aim of the thermal treatment is to oxidise the metals and obtain an oxide with low leachability. Inactive trials were performed, first using small amount of magnesium tape followed by using Cerium instead of uranium, to check the ability of controlling the process. After the process had been deemed safe the next step was to test the process first with metallic uranium and thereafter with magnesium thorium alloy. The first results show that the oxidation process can be totally controlled and safe. The results show that the metals are oxidised and no longer reactive and can in principle be disposed of. The test will continue and further results will be reported. (authors)

  16. Waste form development/test: polyethylene and modified sulfur cement solidification of LLW

    International Nuclear Information System (INIS)

    Kalb, P.D.; Colombo, P.

    1984-01-01

    Low-density polyethylene and a modified sulfur cement have been selected for further study as new potential solidification agents for LLW streams including dry evaporator concentrates, incinerator ash and ion exchange resins. These binder materials are thermoplastic and are processed by use of a screw-type extruder or an air-powered dual action mixer. The studies have utilized ''bench scale'' systems with sufficient capacity to enable reliable determination of the effects of processing parameters such as temperature requirements, feed rates, mixing methods, waste pretreatment and solidification kinetics. Maximum waste loadings have been obtained for polyethylene by use of the screw extruder under controlled conditions. The modified sulfur cement system is processed more effectively by the dual action mixer due to its very low melt viscosity. Simulated waste forms produced using acceptable formulations, are being tested to develop a data base of relevant waste form properties. Several tests underway include immersion, leaching, compressive strength and thermal stability

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

    International Nuclear Information System (INIS)

    Gilliam, T.M.; Hutchins, D.A.; Chodak, P. III

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

  18. 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.).

  19. Hanford Waste Vitrification Plant Quality Assurance Program description for defense high-level waste form development and qualification

    International Nuclear Information System (INIS)

    Hand, R.L.

    1992-01-01

    This document describes the quality assurance (QA) program of the Hanford Waste Vitrification Plant (HWVP) Project. The purpose of the QA program is to control project activities in such a manner as to achieve the mission of the HWVP Project in a safe and reliable manner. A major aspect of the HWVP Project QA program is the control of activities that relate to high-level waste (HLW) form development and qualification. This document describes the program and planned actions the Westinghouse Hanford Company (Westinghouse Hanford) will implement to demonstrate and ensure that the HWVP Project meets the US Department of Energy (DOE) and ASME regulations. The actions for meeting the requirements of the Waste Acceptance Preliminary Specifications (WAPS) will be implemented under the HWVP product qualification program with the objective of ensuring that the HWVP and its processes comply with the WAPS established by the federal repository

  20. 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

  1. 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.

  2. Investigations of the performance of solidified high-level nuclear waste forms

    International Nuclear Information System (INIS)

    Cheung, H.

    1979-01-01

    The Lawrence Livermore Laboratory has been providing technical support to the Nuclear Regulatory Commission in the development of regulations, regulatory guides, and branch technical positions for the management of nuclear wastes. Studies of solidified high-level waste during the period from 1976 to 1978, when work was terminated because of shifting of national emphasis onto spent fuel disposal, are presented in this report. The problem of management, i.e., handling, generation, and disposal of solidified high-level waste derived from operation of commercial light-water reactors, requires a thorough systematic solution to protect health and ensure safety. A definition of the problem by describing the components of the waste management system is given: the waste form, the containers, storage and transportation appurtenances, handling equipment, the repository surface and underground facilities, the repository site, and the operations. A systems analysis methodology to assess the hazards of waste management is developed. Data on accident probabilities, waste form characteristics, and geological and hydrological properties of potential repository sites are compiled. A wide range of management scenarios are generated. Limited sensitivity and uncertainty analyses are performed. On the basis of available information, preliminary investigations showed that transportation and interim storage are of most concern. Also identified are areas needing further study: transportation data base, thermal and seismic aspects of interim storage, human factors, geochemical transport of radionuclides, and ground water composition, among others. In addition to the technical solution of the problems, brief consideration to historical and socioeconomic aspects is also given

  3. Reduction in nitrate leaching from a concrete waste form by using waterproofing admixture additions

    International Nuclear Information System (INIS)

    Kass, M.D.; Lefort, H.G.

    1991-01-01

    Admixture additions of active lime and tricalcium aluminate reduced leach rate about 20% during a seven-day test of a concrete waste form containing a nitrate bearing, aqueous waste. CaO was 5% more effective and, in SEM examination, appeared to hydrate to CaOH which expanded in the concrete pores and blocked water movement. Over time, this compound appeared to dissolve slowly in the water, and the barrier layer it produced slowly receded into the specimen interior

  4. 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.

  5. 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...

  6. Property and process correlations for iron-enriched basalt waste forms

    International Nuclear Information System (INIS)

    Grandy, J.D.; Eddy, T.L.; Anderson, G.L.

    1993-02-01

    Correlations of thermodynamic properties and process parameters of high-temperature slag for a range of compositions of iron-enriched basalt are presented. The quantification of the properties of this complex mixture can assist in the design and monitoring of high-temperature melting systems for the treatment of radioactive and hazardous wastes at the Idaho National Engineering Laboratory. The buried and stored wastes at the INEL Radioactive Waste Management Complex have a similar composition to iron-enriched basalt after oxidation of organics. The properties correlated are the viscosity, electrical conductivity, refractory corrosion, and recrystallization temperature. The correlations are expressed as a function of input waste-soil mixture composition, alkali concentration, and slag temperature. An application to determine the effect of alkali flux on slag temperature, leach rate, and volume reduction is presented. Though the correlations are for mixtures of soil and waste with average transuranic-contaminated waste compositions, it appears that good approximations for other waste streams and glass-ceramic waste forms can be obtained because of similarities in composition

  7. Erosion of magnesium potassium phosphate ceramic waste forms

    International Nuclear Information System (INIS)

    Goretta, K. C.

    1998-01-01

    Phosphate-based chemically bonded ceramics were formed from magnesium potassium phosphate (MKP) binder and either industrial fly ash or steel slag. The resulting ceramics were subjected to solid-particle erosion by a stream of either angular Al 2 O 3 particles or rounded SiO 2 sand. Particle impact angles were 30 or 90degree and the impact velocity was 50 m/s. Steady-state erosion rates, measured as mass lost from a specimen per mass of impacting particle, were dependent on impact angle and on erodent particle size and shape. Material was lost by a combination of fracture mechanisms. Evolution of H 2 O from the MKP phase appeared to contribute significantly to the material loss

  8. Erosion of magnesium potassium phosphate ceramic waste forms.

    Energy Technology Data Exchange (ETDEWEB)

    Goretta, K. C.

    1998-11-20

    Phosphate-based chemically bonded ceramics were formed from magnesium potassium phosphate (MKP) binder and either industrial fly ash or steel slag. The resulting ceramics were subjected to solid-particle erosion by a stream of either angular Al{sub 2}O{sub 3} particles or rounded SiO{sub 2} sand. Particle impact angles were 30 or 90{degree} and the impact velocity was 50 m/s. Steady-state erosion rates, measured as mass lost from a specimen per mass of impacting particle, were dependent on impact angle and on erodent particle size and shape. Material was lost by a combination of fracture mechanisms. Evolution of H{sub 2}O from the MKP phase appeared to contribute significantly to the material loss.

  9. 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.

  10. 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.

  11. 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.

  12. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 2. Commercial waste forms, packaging and projections for preconceptual repository design studies

    International Nuclear Information System (INIS)

    1978-04-01

    This volume, Y/OWI/TM-36/2, ''Commercial Waste Forms, Packaging and Projections for Preconceptual Repository Design Studies,'' 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 provides 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 volume contains the data base for waste forms, packages, and projections from the commercial waste defined by the Office of Waste Isolation in ''Nuclear Waste Projections and Source Term Data for FY 1977,'' Y/OWI/TM-34. Also, as an alternative data base for repository design and analysis, waste forms, packages, and projections for commercial waste defined by Battelle Pacific Northwest Laboratory (BPNL) have been included. This data base consists of a reference case for use in the alternative design study and a definition of combustible wastes for use in mine fire and hydrogen generation analyses

  13. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 2. Commercial waste forms, packaging and projections for preconceptual repository design studies

    Energy Technology Data Exchange (ETDEWEB)

    1978-04-01

    This volume, Y/OWI/TM-36/2, ''Commercial Waste Forms, Packaging and Projections for Preconceptual Repository Design Studies,'' 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 provides 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 volume contains the data base for waste forms, packages, and projections from the commercial waste defined by the Office of Waste Isolation in ''Nuclear Waste Projections and Source Term Data for FY 1977,'' Y/OWI/TM-34. Also, as an alternative data base for repository design and analysis, waste forms, packages, and projections for commercial waste defined by Battelle Pacific Northwest Laboratory (BPNL) have been included. This data base consists of a reference case for use in the alternative design study and a definition of combustible wastes for use in mine fire and hydrogen generation analyses.

  14. 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.

  15. 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.

  16. Corrosion behaviors of a glass-bonded sodalite ceramic waste form and its constituents

    International Nuclear Information System (INIS)

    Lewis, M. A.; Ebert, W. L.; Morss, L.

    1999-01-01

    A ceramic waste form (CWF) of glass bonded sodalite is being developed as a waste form for the long-term immobilization of fission products and transuranic elements from the U.S. Department of Energy's activities on spent nuclear fuel conditioning. A durable waste form was prepared by hot isostatic pressing (HIP) a mixture of salt-loaded zeolite powders and glass frit. During HIP the zeolite is converted to sodalite, and the resultant CWF is been completed for durations of up to 182 days. Four dissolution modes were identified: dissolution of free salt, dissolution of the aluminosilicate matrix of sodalite and the accompanying dissolution of occluded salt, dissolution of the boroaluminosilicate matrix of the glass, and ion exchange. Synergies inherent to the CWF were identified by comparing the results of the tests with pure glass and sodalite with those of the composite CWF

  17. Tracing source and migration of Pb during waste incineration using stable Pb isotopes

    International Nuclear Information System (INIS)

    Li, Yang; Zhang, Hua; Shao, Li-Ming; He, Pin-Jing

    2017-01-01

    Highlights: • The migration of Pb during waste incineration was investigated using Pb isotopes. • Source tracing of Pb during incineration by isotopic technology was feasible. • Contributions of MSW components were measured to trace Pb sources quantitatively. • Isotopic technology helps understand the migration of Pb during thermal treatment. - Abstract: Emission of Pb is a significant environmental concern during solid waste incineration. To target Pb emission control strategies effectively, the major sources of Pb in the waste incineration byproducts must be traced and quantified. However, identifying the migration of Pb in each waste component is difficult because of the heterogeneity of the waste. This study used a laboratory-scale incinerator to simulate the incineration of municipal solid waste (MSW). The Pb isotope ratios of the major waste components ( 207 Pb/ 206 Pb = 0.8550–0.8627 and 208 Pb/ 206 Pb = 2.0957–2.1131) and their incineration byproducts were measured to trace sources and quantify the Pb contribution of each component to incineration byproducts. As the proportions of food waste (FW), newspaper (NP), and polyethylene bag (PE) in the artificial MSW changed, the contribution ratios of FW and PE to Pb in fly ash changed accordingly, ranging from 31.2% to 50.6% and from 35.0% to 41.8%, respectively. The replacement of PE by PVC significantly increased the partitioning and migration ratio of Pb. The use of Pb isotope ratios as a quantitative tool for tracing Pb from raw waste to incineration byproducts is a feasible means for improving Pb pollution control.

  18. Tracing source and migration of Pb during waste incineration using stable Pb isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yang [State Key Laboratory of Pollution Control and Resources Reuse, Tongji University, 1239 Siping Road, Shanghai 200092 (China); Institute of Waste Treatment and Reclamation, Tongji University, 1239 Siping Road, Shanghai 200092 (China); Zhang, Hua, E-mail: zhanghua_tj@tongji.edu.cn [State Key Laboratory of Pollution Control and Resources Reuse, Tongji University, 1239 Siping Road, Shanghai 200092 (China); Institute of Waste Treatment and Reclamation, Tongji University, 1239 Siping Road, Shanghai 200092 (China); Shao, Li-Ming; He, Pin-Jing [Institute of Waste Treatment and Reclamation, Tongji University, 1239 Siping Road, Shanghai 200092 (China); Research and Training Center on Rural Waste Management, Ministry of Housing and Urban-Rural Development of P.R. China, 1239 Siping Road, Shanghai 200092 (China)

    2017-04-05

    Highlights: • The migration of Pb during waste incineration was investigated using Pb isotopes. • Source tracing of Pb during incineration by isotopic technology was feasible. • Contributions of MSW components were measured to trace Pb sources quantitatively. • Isotopic technology helps understand the migration of Pb during thermal treatment. - Abstract: Emission of Pb is a significant environmental concern during solid waste incineration. To target Pb emission control strategies effectively, the major sources of Pb in the waste incineration byproducts must be traced and quantified. However, identifying the migration of Pb in each waste component is difficult because of the heterogeneity of the waste. This study used a laboratory-scale incinerator to simulate the incineration of municipal solid waste (MSW). The Pb isotope ratios of the major waste components ({sup 207}Pb/{sup 206}Pb = 0.8550–0.8627 and {sup 208}Pb/{sup 206}Pb = 2.0957–2.1131) and their incineration byproducts were measured to trace sources and quantify the Pb contribution of each component to incineration byproducts. As the proportions of food waste (FW), newspaper (NP), and polyethylene bag (PE) in the artificial MSW changed, the contribution ratios of FW and PE to Pb in fly ash changed accordingly, ranging from 31.2% to 50.6% and from 35.0% to 41.8%, respectively. The replacement of PE by PVC significantly increased the partitioning and migration ratio of Pb. The use of Pb isotope ratios as a quantitative tool for tracing Pb from raw waste to incineration byproducts is a feasible means for improving Pb pollution control.

  19. 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.

  20. 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.

  1. Metal waste forms from the electrometallurgical treatment of spent nuclear fuel

    International Nuclear Information System (INIS)

    Abraham, D.P.; McDeavitt, S.M.; Park, J.

    1996-01-01

    Stainless steel-zirconium alloys are being developed for the disposal of radioactive metal isotopes isolated using an electrometallurgical treatment technique to treat spent nuclear fuel. The nominal waste forms are stainless steel-15 wt% zirconium alloy and zirconium-8 wt% stainless steel alloy. These alloys are generated in yttria crucibles by melting the starting materials at 1,600 C under an argon atmosphere. This paper discusses the microstructures, corrosion and mechanical test results, and thermophysical properties of the metal waste form alloys

  2. 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.

  3. Leaching of radioactive waste forms under saturated and unsaturated flow conditions

    International Nuclear Information System (INIS)

    Petelka, M.F.

    1987-01-01

    To predict the environmental impact of shallow land burial sites for radioactive waste, the mobilization and migration of waste nuclides must be estimated. The theoretical understanding that in potential leaching mechanisms leach-rate variations may arise from changes in both moisture content and volumetric flow rate was tested in column flow leach experiments using labeled vermiculite particles as a simulated waste form. As far as possible, conditions of flow rate and solution ion concentration were chosen to roughly approximate expected field conditions. A modified pressure-plate apparatus was developed, tested, and found suitable for the production of steady-state unsaturated conditions with leachate flow. Water content was determined using the gamma-ray attenuation method. The effects of several parameters on leaching were studied, including moisture content and pore velocity. Pore velocity effects were found to be negligible. It was found that the leach rate depends on the fraction of the exposed waste surface that is wetted and varies with the mobile water content in a non-linear fashion. The experimental results indicate that the release rate of radionuclides placed within a properly sited low-level waste disposal site may be two to three times smaller than that predicted assuming saturated conditions. This study was performed using a homogeneous fine-grained synthetic waste form, at room temperature, with a near neutral pH leachant and oxidizing conditions

  4. Measurements of Mercury Released From Solidified/Stabilized Waste Forms-FY2002

    International Nuclear Information System (INIS)

    Mattus, C.H.

    2003-01-01

    This report covers work performed during FY 2002 in support of treatment demonstrations conducted for the U.S. Department of Energy (DOE) Mixed Waste Focus Area (MWFA) Mercury Working Group. To comply with the requirements of the Resource Conservation and Recovery Act, as implemented by the U.S. Environmental Protection Agency (EPA), DOE must use one of the following procedures for mixed low-level radioactive wastes containing mercury at levels above 260 ppm: a retorting/roasting treatment or (if the wastes also contain organics) an incineration treatment. The recovered radioactively contaminated mercury must then be treated by an amalgamation process prior to disposal. The DOE MWFA Mercury Working Group is working with EPA to determine whether some alternative processes could be used to treat these types of waste directly, thereby avoiding a costly recovery step for DOE. In previous years, demonstrations were performed in which commercial vendors applied their technologies for the treatment of radiologically contaminated elemental mercury as well as radiologically contaminated and mercury-contaminated waste soils from Brookhaven National Laboratory. The test results for mercury release in the headspace were reported in two reports, ''Measurements of Mercury Released from Amalgams and Sulfide Compounds'' (ORNL/TM-13728) and ''Measurements of Mercury Released from Solidified/Stabilized Waste Forms'' (ORNL/TM-2001/17). The current work did not use a real waste; a surrogate sludge had been prepared and used in the testing in an effort to understand the consequences of mercury speciation on mercury release

  5. Leaching behavior of a simulated bituminized radioactive waste form under deep geological conditions

    International Nuclear Information System (INIS)

    Nakayama, Shinichi; Iida, Yoshihisa; Nagano, Tetsushi; Akimoto, Toshiyuki

    2003-01-01

    The leaching behavior of a simulated bituminized waste form was studied to acquire data for the performance assessment of the geologic disposal of bituminized radioactive waste. Laboratory-scale leaching tests were performed for radioactive and non-radioactive waste specimens simulating bituminized waste of a French reprocessing company, COGEMA. The simulated waste was contacted with deionized water, an alkaline solution (0.03-mol/l KOH), and a saline solution (0.5-mol/l KCl) under atmospheric and anoxic conditions. The concentrations of Na, Ba, Cs, Sr, Np, Pu, NO 3 , SO 4 and I in the leachates were determined. Swelling of the bituminized waste progressed in deionized water and KOH. The release of the soluble components, Na and Cs, was enhanced by the swelling, and considered to be diffusion-controlled in the swelled layers of the specimens. The release of sparingly soluble components such as Ba and Np was solubility-limited in addition to the progression of leaching. Neptunium, a redox-sensitive element, showed a distinct difference in release between anoxic and atmospheric conditions. The elemental release from the bituminized waste specimens leached in the KCl was very low, which is likely due to the suppression of swelling of the specimens at high ionic strength. (author)

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

    International Nuclear Information System (INIS)

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

    1979-12-01

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

  7. Comparisons between longtime stable environmentally toxic substances and nuclear waste - a literature study

    International Nuclear Information System (INIS)

    Persson, L.

    1988-01-01

    The chemical and radiological risks from nuclear wastes can be compared as they both have genotoxic effects. Comments are given to the problem of comparing risks from different genotoxic substances. (97 refs.) (L.E.)

  8. 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

  9. The promyelocytic leukemia gene product (PML) forms stable complexes with the retinoblastoma protein

    DEFF Research Database (Denmark)

    Alcalay, M; Tomassoni, L; Colombo, E

    1998-01-01

    by the expression of PML-RAR alpha. We report that PML colocalizes with the nonphosphorylated fraction of the retinoblastoma protein (pRB) within nuclear bodies and that pRB is delocalized by PML-RAR alpha expression. Both PML and PML-RAR alpha form complexes with the nonphosphorylated form of pRB in vivo...

  10. 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.

  11. Secondary phases formed during nuclear waste glass-water interactions: Thermodynamic and derived properties

    International Nuclear Information System (INIS)

    McKenzie, W.F.

    1992-08-01

    The thermodynamic properties of secondary phases observed to form during nuclear waste glass-water interactions are of particular interest as it is with the application of these properties together with the thermodynamic properties of other solid phases, fluid phases, and aqueous species that one may predict the environmental consequences of introducing radionuclides contained in the glass into groundwater at a high-level nuclear waste repository. The validation of these predicted consequences can be obtained from laboratory experiments and field observations at natural analogue sites. The purpose of this report is to update and expand the previous compilation (McKenzie, 1991) of thermodynamic data retrieved from the literature and/or estimated for secondary phases observed to form (and candidate phases from observed chemical compositions) during nuclear waste glass-water interactions. In addition, this report includes provisionally recommended thermodynamic data of secondary phases

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  13. Characterization and testing of a 238Pu loaded ceramic waste form

    International Nuclear Information System (INIS)

    Johnson, S. G.

    1998-01-01

    This paper will describe the preparation and progress of the effort at Argonne National Laboratory-West to produce ceramic waste forms loaded with 238 Pu. The purpose of this study is to determine the extent of damage, if any, that alpha decay events will play over time to the ceramic waste form under development at Argonne. The ceramic waste form is glass-bonded sodalite. The sodalite is utilized to encapsulate the fission products and transuranics which are present in a chloride salt matrix which results from a spent fuel conditioning process. 238 Pu possesses approximately 250 times the specific activity of 239 Pu and thus allows for a much shorter time frame to address the issue. In preparation for production of 238 Pu loaded waste forms 239 Pu loaded samples were produced. Data is presented for samples produced with typical reactor grade plutonium. X-ray diffraction, scanning electron micrographs and durability test results will be presented. The ramifications for the production of the 238 Pu loaded samples will be discussed

  14. Characterization and testing of a {sup 238}Pu loaded ceramic waste form.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, S. G.

    1998-04-24

    This paper will describe the preparation and progress of the effort at Argonne National Laboratory-West to produce ceramic waste forms loaded with {sup 238}Pu. The purpose of this study is to determine the extent of damage, if any, that alpha decay events will play over time to the ceramic waste form under development at Argonne. The ceramic waste form is glass-bonded sodalite. The sodalite is utilized to encapsulate the fission products and transuranics which are present in a chloride salt matrix which results from a spent fuel conditioning process. {sup 238}Pu possesses approximately 250 times the specific activity of {sup 239}Pu and thus allows for a much shorter time frame to address the issue. In preparation for production of {sup 238}Pu loaded waste forms {sup 239}Pu loaded samples were produced. Data is presented for samples produced with typical reactor grade plutonium. X-ray diffraction, scanning electron micrographs and durability test results will be presented. The ramifications for the production of the {sup 238}Pu loaded samples will be discussed.

  15. Neutron activation analysis of alternative waste forms at the Savannah River Laboratory

    International Nuclear Information System (INIS)

    Johns, R.A.

    1981-01-01

    A remotely controlled system for neutron activation of candidate high-level waste (HLW) isolation forms was built by the Savannah River Laboratory at a Savannah River Plant reactor. With this system, samples can be irradiated for up to 24 hours and transferred through pneumatic tubing to a shielded repository unitl their activity is low enough for them to be handled in a radiobench. The principal use of the system is to support the Alternative Waste Forms Leach Testing (AWFLT) Program in which the comparative leachability of the various waste forms will be determined. The experimental method used in this work is based on neutron activation analysis techniques. Neutron irradiation of the solid waste form containing simulated HLW sludge activates elements in the sample. After suitable leaching of the solid matrix in standard solutions, the leachate and solid are assayed for gamma-emitting nuclides. From these measurements, the fraction of a specific element leached can be determined al half-lives with experimental ones, over a range of 24 orders of magnitude was obtained. This is a strong argument that the alpha decay could be considered a fission process with very high mass asymmetry and charge density asymmetry

  16. 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.

  17. 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.

  18. Effect of aluminum and silicon reactants and HIP soak time on characteristics of glass-ceramic waste forms

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1993-04-01

    The high level liquid waste (HLLW) from nuclear fuel reprocessing is being calcined into solid granules and being stored onsite at the Idaho Chemical Processing Plant (ICPP) since 1963. Final disposal of the calcined waste in a geologic repository requires further consolidation of the calcine in to a solid waste form. One of the solid waste forms being considered for immobilization of the ICPP calcines is the glass-ceramic. The glass-ceramic waste form is a promising option because it can potentially reduce the calcined high level waste (HLW) volume significantly compared to glass waste forms while maintaining similar leach rates. Based on technical evaluations, and laboratory and pilot plant mockup tests, the Environmental Protection Agency (EPA) believes that the glass-ceramic process is more efficient than the glass process for ICPP calcine waste forms. The EPA has determined that the glass-ceramic waste form technology is an acceptable technology to meet the Best Demonstrated Acceptable Technology (BDAT) for ICPP HLW calcine. In this progress report, the impact of aluminum and silicon reactants and HIP soak time on leach rates, microstructure and phase composition of glass-ceramic waste forms are discussed

  19. 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.

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

    International Nuclear Information System (INIS)

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

    1988-08-01

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

  1. Research on the skirt tail explosively formed projectile stable shaping technology

    Science.gov (United States)

    Li, W.; Wang, X.; Li, W.; Chen, K.

    2016-09-01

    The effect of the arc-cone liner configuration parameters on explosively formed projectile (EFP) flight stability is studied experimentally. The effect of the liner edge structure on the EFP formation is computed numerically. The results show that an EFP formed by a liner whose thickness is 0.046 times the charge caliber has improved flight stability and can perforate a steel armor 0.5 times the charge caliber thick in the case of large stand-off distances. A good tail skirt EFP can be formed by choosing appropriate parameters: the liner-to-charge diameter ratio in the interval 0.96-0.98, the liner edge thickness equal to 0.0020-0.0025 times the charge caliber, and the liner edge chamfer of 45-50°.

  2. Materials Characterization Center. Second workshop on irradiation effects in nuclear waste forms. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Weber, W.J.; Turcotte, R.P.

    1982-01-01

    The purpose of this second workshop on irradiations effects was to continue the discussions initiated at the first workshop and to obtain guidance for the Materials Characterization Center in developing test methods. The following major conclusions were reached: Ion or neutron irradiations are not substitutes for the actinide-doping technique, as described by the MCC-6 Method for Preparation and Characterization of Actinide-Doped Waste Forms, in the final evaluation of any waste form with respect to the radiation effects from actinide decay. Ion or neutron irradiations may be useful for screening tests or more fundamental studies. The use of these simulation techniques as screening tests for actinide decay requires that a correlation between ion or neutron irradiations and actinide decay be established. Such a correlation has not yet been established and experimental programs in this area are highly recommended. There is a need for more fundamental studies on dose-rate effects, temperature dependence, and the nature and importance of alpha-particle effects relative to the recoil nucleus in actinide decay. There are insufficient data presently available to evaluate the potential for damage from ionizing radiation in nuclear waste forms. No additional test methods were recommended for using ion or neutron irradiations to simulate actinide decay or for testing ionization damage in nuclear waste forms. It was recognized that additional test methods may be required and developed as more data become available. An American Society for Testing and Materials (ASTM) Task Group on the Simulation of Radiation Effects in Nuclear Waste Forms (E 10.08.03) was organized to act as a continuing vehicle for discussions and development of procedures, particularly with regard to ion irradiations.

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

    International Nuclear Information System (INIS)

    1983-10-01

    The conceptual waste package interim product specifications and data requirements presented are applicable to the reference glass composition described in PNL-3838 and carbon steel canister described in ONWI-438. They provide preliminary numerical values for the commercial 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. 13 references, 1 figure

  4. STUDY OF STABLE NITROGEN FORMS IN NATURAL SURFACE WATERS IN THE PRESENCE OF MINERAL SUBSTRATES

    Directory of Open Access Journals (Sweden)

    Petru Spataru

    2015-12-01

    Full Text Available The influence of substrates on the oxidation of reduced toxic forms of nitrogen in river water was investigated by laboratory modelling. Granite and expended clay accelerate the oxidation of ammonium and nitrite ions from 2 to 4 times. The presence of calcium carbonate in water hinders the oxidation of nitrogen in the polluted water.

  5. Design of a stable steam reforming catalyst - A promising route to sustainable hydrogen form biomass oxygenates

    NARCIS (Netherlands)

    Matas Güell, B.; Babych, Igor V.; Nichols, K.P.F.; Gardeniers, Johannes G.E.; Lefferts, Leonardus; Seshan, Kulathuiyer

    2009-01-01

    The influence of the support and the presence of oxygen were investigated in the steam reforming of acetic acid, a bio-oil model compound, over Pt/ZrO2 and Pt/CeO2 catalysts. In the absence of oxygen, all catalysts suffered from deactivation. Acetone, formed via condensation/dehydration of acetic

  6. Glass Waste Forms for Oak Ridge Tank Wastes: Fiscal Year 1998 Report for Task Plan SR-16WT-31, Task B

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, M.K.

    1999-05-10

    Using ORNL information on the characterization of the tank waste sludges, SRTC performed extensive bench-scale vitrification studies using simulants. Several glass systems were tested to ensure the optimum glass composition (based on the glass liquidus temperature, viscosity and durability) is determined. This optimum composition will balance waste loading, melt temperature, waste form performance and disposal requirements. By optimizing the glass composition, a cost savings can be realized during vitrification of the waste. The preferred glass formulation was selected from the bench-scale studies and recommended to ORNL for further testing with samples of actual OR waste tank sludges.

  7. Atmospherically stable nanoscale zero-valent iron particles formed under controlled air contact: characteristics and reactivity.

    Science.gov (United States)

    Kim, Hong-Seok; Ahn, Jun-Young; Hwang, Kyung-Yup; Kim, Il-Kyu; Hwang, Inseong

    2010-03-01

    Atmospherically stable NZVI (nanoscale zero-valent iron) particles were produced by modifying shell layers of Fe(H2) NZVI particles (RNIP-10DS) by using a controlled air contact method. Shell-modified NZVI particles were resistant to rapid aerial oxidation and were shown to have TCE degradation rate constants that were equivalent to 78% of those of pristine NZVI particles. Fe(H2) NZVI particles that were vigorously contacted with air (rapidly oxidized) showed a substantially compromised reactivity. Aging of shell-modified particles in water for one day resulted in a rate increase of 54%, implying that depassivation of the shell would play an important role in enhancing reactivity. Aging of shell-modified particles in air led to rate decreases by 14% and 46% in cases of one week and two months of aging, respectively. A series of instrumental analyses using transmission electron microscopy, X-ray diffractography, X-ray photoelectron spectroscopy, and X-ray absorption near-edge structure showed that the shells of modified NZVI particles primarily consisted of magnetite (Fe(3)O(4)). Analyses also implied that the new magnetite layer produced during shell modification was protective against shell passivation. Aging of shell-modified particles in water yielded another major mineral phase, goethite (alpha-FeOOH), whereas aging in air produced additional shell phases such as wustite (FeO), hematite (alpha-Fe(2)O(3)), and maghemite (gamma-Fe(2)O(3)).

  8. Solution exchange corrosion testing with the glass-zeolite ceramic waste form in demineralized water at 900C

    International Nuclear Information System (INIS)

    Simpson, L. J.

    1998-01-01

    A ceramic waste form of glass-bonded zeolite is being developed for the long-term disposition of fission products and transuranic elements in wastes from the U.S. Department of Energy's spent nuclear fuel conditioning activities. Solution exchange corrosion tests were performed on the ceramic waste form and its potential base constituents of glass, zeolite 5A, and sodalite as part of an effort to qualify the ceramic waste form for acceptance into the Civilian Radioactive Waste Management System. Solution exchange tests were performed at 90 C by replacing 80 to 90% of the leachate with fresh demineralized water after set time intervals. The results from these tests provide information about corrosion mechanisms and the ability of the ceramic waste form and its constituent materials to retain waste components. The results from solution exchange tests indicate that radionuclides will be preferentially retained in the zeolites without the glass matrix and in the ceramic waste form, with respect to cations like Li, K, and Na. Release results have been compared for simulated waste from candidate ceramic waste forms with zeolite 5A and its constituent materials to determine the corrosion behavior of each component

  9. Step-function increase of significance of waste form in RWM strategy

    International Nuclear Information System (INIS)

    Roy, R.

    1979-01-01

    Radioactive waste management on the technical side is composed of four subsystems: storage, radionuclide immobilization, isolation, and post-emplacement. Thus far, emphasis has been on the isolation subsystem, with the assumption that all the radionuclides placed in the geological formation would be released. Findings by a committee of the National Academy of Science on the types of solid forms for radioactive wastes are summarized. The most important research findings of the last two or three years are evaluated, findings which impinge on the role of the newly developed radionuclide immobilization subsystems. 24 refs

  10. TRU waste form studies with special reference to iron-enriched basalt: 1980. Annual report

    International Nuclear Information System (INIS)

    Flinn, J.E.; Henslee, S.P.; Kelsey, P.V. Jr.

    1981-06-01

    Material studies were performed on iron-enriched basalt (IEB) as a waste form containment medium for transuranic wastes. Specimens from laboratory scale, as well as large scale melts, were used in the evaluation. The studies included melting and casting, slag-refractory interaction, slag fruit assessments, volatility of sodium salts from IEB melts, chemical and structure homogeneity, metallic dissolution tests, physical properties, and devitrification associated with the development of mineral phases. In addition, durability tests, which included leaching and mechanical behavior, were performed

  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. 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.

  13. Investigation of activity release from bituminized intermediate-level waste forms under thermal stresses

    International Nuclear Information System (INIS)

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

    1983-01-01

    To determine the consequences of a fire during fabrication, intermediate storage and transport of bituminized NaNO 3 waste forms, the fractions of plutonium released from the waste forms were assessed. For this purpose, laboratory tests were made with PuO 2 -containing specimens as well as a field test with specimens containing Eu 2 O 3 . By the evaluation of plutonium release in the laboratory and by the determination of the total sodium release and the relative Eu/Na release in the field tests the plutonium release can be deduced from full-scale specimens. The results show that for bituminized waste forms with high NaNO 3 contents (approx. 36 wt%) the average plutonium release obtained in laboratory testing is 15%. In the field tests (IAEA fire test conditions) an average Eu release of 8% was found. These results justify the statement that also for waste forms in open 175 L drum inserts a maximum plutonium release of about 15% can be expected. From the time-dependence of Eu/Na release in the field tests an induction period of 15-20 minutes between the start of testing and the first Na/Eu release can be derived. The maximum differential Na/Eu release occurs after a test period of 45 to 60 minutes duration and after 90 to 105 minutes (tests K2 and K4, respectively); after that time also the highest temperatures in the products are measured. The release values were determined for products in open 175 L drum inserts which in this form are not eligible for intermediate and ultimate storage. For bituminized waste forms in concrete packages (lost concrete shieldings) a delayed increase in temperature to only 70-80 deg. C takes place (4-5 hours after extinction of the fire) if the fire lasts 45 minutes. The concrete package remains intact under test conditions. This means that activity release from bituminized waste forms packaged in this way can be ruled out in the case under consideration. (author)

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

    International Nuclear Information System (INIS)

    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.''

  15. Capric-myristic acid/expanded perlite composite as form-stable phase change material for latent heat thermal energy storage

    Energy Technology Data Exchange (ETDEWEB)

    Karaipekli, Ali; Sari, Ahmet [Department of Chemistry, Gaziosmanpasa University, 60240 Tokat (Turkey)

    2008-12-15

    The aim of this study is to prepare a novel form-stable phase change material (PCM) for latent heat thermal energy storage (LHTES) in buildings. A eutectic mixture of capric acid (CA) and myristic acid (MA) is incorporated with expanded perlite (EP). Thermal properties, thermal reliability, and thermal conductivity of the form-stable composite PCM are determined. The maximum CA-MA absorption of EP was found to be 55 wt% without melted PCM seepage from the composite, and therefore this mixture was described as a form-stable composite. The form-stable composite PCM was characterized using the FT-IR spectroscopy method. The melting and freezing temperatures and latent heats of form-stable composite PCM were measured using DSC analysis. Thermal cycling test of the form-stable composite PCM indicated good thermal reliability in terms of changes in thermal properties after 5000 thermal cycling. The thermal conductivity of the form-stable CA-MA/EP composite PCM was increased about 58% by adding 10 wt% expanded graphite (EG). The form-stable CA-MA/EP/EG composite PCM was considered as an effective LHTES material in the building energy conservation due to suitable phase change temperatures, high latent capacities, good thermal reliability, and good thermal conductivity. (author)

  16. Pu(V) as the stable form of oxidized plutonium in natural waters

    International Nuclear Information System (INIS)

    Orlandini, K.A.; Penrose, W.R.; Nelson, D.M.

    1986-01-01

    This work presents analytical evidence supporting the proposition that Pu(V) is the sole or predominant form of oxidized plutonium in natural waters. Two independent methods, the selective adsorption of Pu(VI) by silica gel, and the somewhat less selective coprecipitation of Pu(V) with calcium carbonate, were developed to separate Pu(V) from Pu(VI). Measurements of ambient plutonium in several natural waters by these methods found only Pu(V). In laboratory tracer studies, Pu(VI) was shown to be highly unstable in dilute bicarbonate solution and in Lake Michigan water, reducing in first-order fashion to Pu(V). (orig.)

  17. 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

  18. Ceramic Single Phase High-Level Nuclear Waste Forms: Hollandite, Perovskite, and Pyrochlore

    Science.gov (United States)

    Vetter, M.; Wang, J.

    2017-12-01

    The lack of viable options for the safe, reliable, and long-term storage of nuclear waste is one of the primary roadblocks of nuclear energy's sustainable future. The method being researched is the incorporation and immobilization of harmful radionuclides (Cs, Sr, Actinides, and Lanthanides) into the structure of glasses and ceramics. Borosilicate glasses are the main waste form that is accepted and used by today's nuclear industry, but they aren't the most efficient in terms of waste loading, and durability is still not fully understood. Synroc-phase ceramics (i.e. hollandite, perovskite, pyrochlore, zirconolite) have many attractive qualities that glass waste forms do not: high waste loading, moderate thermal expansion and conductivity, high chemical durability, and high radiation stability. The only downside to ceramics is that they are more complex to process than glass. New compositions can be discovered by using an Artificial Neural Network (ANN) to have more options to optimize the composition, loading for performance by analyzing the non-linear relationships between ionic radii, electronegativity, channel size, and a mineral's ability to incorporate radionuclides into its structure. Cesium can be incorporated into hollandite's A-site, while pyrochlore and perovskite can incorporate actinides and lanthanides into their A-site. The ANN is used to predict new compositions based on hollandite's channel size, as well as the A-O bond distances of pyrochlore and perovskite, and determine which ions can be incorporated. These new compositions will provide more options for more experiments to potentially improve chemical and thermodynamic properties, as well as increased waste loading capabilities.

  19. Measurements of Mercury Released From Solidified/Stabilized Waste Forms-FY2002

    Energy Technology Data Exchange (ETDEWEB)

    Mattus, C.H.

    2003-02-17

    This report covers work performed during FY 2002 in support of treatment demonstrations conducted for the U.S. Department of Energy (DOE) Mixed Waste Focus Area (MWFA) Mercury Working Group. To comply with the requirements of the Resource Conservation and Recovery Act, as implemented by the U.S. Environmental Protection Agency (EPA), DOE must use one of the following procedures for mixed low-level radioactive wastes containing mercury at levels above 260 ppm: a retorting/roasting treatment or (if the wastes also contain organics) an incineration treatment. The recovered radioactively contaminated mercury must then be treated by an amalgamation process prior to disposal. The DOE MWFA Mercury Working Group is working with EPA to determine whether some alternative processes could be used to treat these types of waste directly, thereby avoiding a costly recovery step for DOE. In previous years, demonstrations were performed in which commercial vendors applied their technologies for the treatment of radiologically contaminated elemental mercury as well as radiologically contaminated and mercury-contaminated waste soils from Brookhaven National Laboratory. The test results for mercury release in the headspace were reported in two reports, ''Measurements of Mercury Released from Amalgams and Sulfide Compounds'' (ORNL/TM-13728) and ''Measurements of Mercury Released from Solidified/Stabilized Waste Forms'' (ORNL/TM-2001/17). The current work did not use a real waste; a surrogate sludge had been prepared and used in the testing in an effort to understand the consequences of mercury speciation on mercury release.

  20. Representations of Political Power Structures by Strategically Stable Game Forms: A Survey

    Directory of Open Access Journals (Sweden)

    Bezalel Peleg

    2017-10-01

    Full Text Available We survey the results on representations of committees and constitutions by game forms that possess some kind of equilibrium strategies for each profile of preferences of the players. The survey is restricted to discrete models, that is, we deal with finitely many players and alternatives. No prior knowledge of social choice is assumed: As far as definitions are concerned, the paper is self-contained. Section 2 supplies the necessary general tools for the rest of the paper. Each definition is followed by a simple (but nontrivial example. In Section 3 we give a complete account of representations of committees (proper and monotonic simple games, by exactly and strongly consistent social choice functions. We start with Peleg’s representations of weak games, and then provide a complete and detailed account of Holzman’s solution of the representation problem for simple games without veto players. In Section 4 we deal with representations of constitutions by game forms. Following Gärdenfors we model a constitution by a monotonic and superadditive effectivity function. We fully characterize the representations for three kinds of equilibrium: Nash equilibrium; acceptable equilibrium (Pareto optimal Nash equilibrium; and strong Nash equilibrium. We conclude in Section 5 with a report on two recent works on representations of constitutions under incomplete information.

  1. Graphene/phase change material nanocomposites: light-driven, reversible electrical resistivity regulation via form-stable phase transitions.

    Science.gov (United States)

    Wang, Yunming; Mi, Hongyi; Zheng, Qifeng; Ma, Zhenqiang; Gong, Shaoqin

    2015-02-04

    Innovative photoresponsive materials are needed to address the complexity of optical control systems. Here, we report a new type of photoresponsive nanomaterial composed of graphene and a form-stable phase change material (PCM) that exhibited a 3 orders of magnitude change in electrical resistivity upon light illumination while retaining its overall original solid form at the macroscopic level. This dramatic change in electrical resistivity also occurred reversibly through the on/off control of light illumination. This was attributed to the reversible phase transition (i.e., melting/recrystallization) behavior of the microscopic crystalline domains present in the form-stable PCM. The reversible phase transition observed in the graphene/PCM nanocomposite was induced by a reversible temperature change through the on/off control of light illumination because graphene can effectively absorb light energy and convert it to thermal energy. In addition, this graphene/PCM nanocomposite also possessed excellent mechanical properties. Such photoresponsive materials have many potential applications, including flexible electronics.

  2. Testing and evaluation of solidified high-level waste forms. Joint annual progress report 1983

    International Nuclear Information System (INIS)

    Malow, G.

    1985-01-01

    A second joint programme of the European Atomic Community was started in 1981 under the indirect action programme (1980-84), Action No 5 'Testing and evaluation of the properties of various potential materials for immobilizing high activity waste'. The overall objective of the research is to test various European potential solidified high-level radioactive waste forms so as to predict their behaviour after disposal. The most important aspect is to produce data to calculate the activity release from the waste products under the attack of various aqueous solutions. The experiments were partly performed under waste repository relevant conditions and partly under simplified conditions for investigating basic activity release mechanisms. The topics of the programme were: (i) studies of basic leaching mechanisms; (ii) studies of hydrothermal leaching and surface attack of waste glasses; (iii) leach test carried out in contact with granite at low water flow rates; (iv) static leach tests with specimen surrounded by canister and backfill materials; (v) specific isotope leach tests in slowly flowing water; (vi) leach test of actinide spiked samples; (vii) leach tests of highly radioactive samples; (viii) leach tests of alpha radiation stability; (ix) studies of mechanical stability; (x) studies of mineral phases as model compounds and phase relations

  3. Characterization of radioactive waste forms. Progress report for 1986 Volume 1

    International Nuclear Information System (INIS)

    Glasser, F.P.; McCulloch, C.

    1988-01-01

    The Council of Ministers of the European Communities adopted the third five-year EC programme of research on radioactive waste management and disposal in March 1985. It was recognized that the inevitable production of radioactive waste required perfecting and demonstrating systems for managing the waste produced by the nuclear industry, ensuring at the various stages the best possible protection of man and the environment. Task 3 of the programme 'evaluation of conditioned waste and qualification of engineered barriers' is subdivided into five sections. This book, in two volumes, is a compilation of reports on the progress achieved in four of the sections during 1986, the first year of the third programme. Volume 1 is concerned with Sections 1 and 5, 'Research on low -and medium- active waste' and 'Quality control methods'. Volume 2 covers Section 2 'HLW form characterization' and Section 3 'Other engineered barriers'. Section 4 'Development of standard test methods' is not included in this edition, as results from an interlaboratory round robin test now in progress will only be available for inclusion for the year 1988

  4. Porphyry-copper ore shells form at stable pressure-temperature fronts within dynamic fluid plumes.

    Science.gov (United States)

    Weis, P; Driesner, T; Heinrich, C A

    2012-12-21

    Porphyry-type ore deposits are major resources of copper and gold, precipitated from fluids expelled by crustal magma chambers. The metals are typically concentrated in confined ore shells within vertically extensive vein networks, formed through hydraulic fracturing of rock by ascending fluids. Numerical modeling shows that dynamic permeability responses to magmatic fluid expulsion can stabilize a front of metal precipitation at the boundary between lithostatically pressured up-flow of hot magmatic fluids and hydrostatically pressured convection of cooler meteoric fluids. The balance between focused heat advection and lateral cooling controls the most important economic characteristics, including size, shape, and ore grade. This self-sustaining process may extend to epithermal gold deposits, venting at active volcanoes, and regions with the potential for geothermal energy production.

  5. Converting Simulated Sodium-bearing Waste into a Single Solid Waste Form by Evaporation: Laboratory- and Pilot-Scale Test Results on Recycling Evaporator Overheads

    International Nuclear Information System (INIS)

    Griffith, D.; D. L. Griffith; R. J. Kirkham; L. G. Olson; S. J. Losinski

    2004-01-01

    Conversion of Idaho National Engineering and Environmental Laboratory radioactive sodium-bearing waste into a single solid waste form by evaporation was demonstrated in both flask-scale and pilot-scale agitated thin film evaporator tests. A sodium-bearing waste simulant was adjusted to represent an evaporator feed in which the acid from the distillate is concentrated, neutralized, and recycled back through the evaporator. The advantage to this flow sheet is that a single remote-handled transuranic waste form is produced in the evaporator bottoms without the generation of any low-level mixed secondary waste. However, use of a recycle flow sheet in sodium-bearing waste evaporation results in a 50% increase in remote-handled transuranic volume in comparison to a non-recycle flow sheet

  6. 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.

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

    International Nuclear Information System (INIS)

    Aguilar, R.

    2003-01-01

    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

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

    International Nuclear Information System (INIS)

    Armantrout, G.A.; Jardine, L.J.

    1996-02-01

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

  9. 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.

  10. Scale up issues involved with the ceramic waste form: ceramic-container interactions and ceramic cracking quantification

    International Nuclear Information System (INIS)

    Bateman, K. J.; DiSanto, T.; Goff, K. M.; Johnson, S. G.; O'Holleran, T.; Riley, W. P. Jr.

    1999-01-01

    Argonne National Laboratory is developing a process for the conditioning of spent nuclear fuel to prepare the material for final disposal. Two waste streams will result from the treatment process, a stainless steel based form and a ceramic based form. The ceramic waste form will be enclosed in a stainless steel container. In order to assess the performance of the ceramic waste form in a repository two factors must be examined, the surface area increases caused by waste form cracking and any ceramic/canister interactions that may release toxic material. The results indicate that the surface area increases are less than the High Level Waste glass and any toxic releases are below regulatory limits

  11. CAG Expansions Are Genetically Stable and Form Nontoxic Aggregates in Cells Lacking Endogenous Polyglutamine Proteins

    Directory of Open Access Journals (Sweden)

    Ashley A. Zurawel

    2016-09-01

    Full Text Available Proteins containing polyglutamine (polyQ regions are found in almost all eukaryotes, albeit with various frequencies. In humans, proteins such as huntingtin (Htt with abnormally expanded polyQ regions cause neurodegenerative diseases such as Huntington’s disease (HD. To study how the presence of endogenous polyQ aggregation modulates polyQ aggregation and toxicity, we expressed polyQ expanded Htt fragments (polyQ Htt in Schizosaccharomyces pombe. In stark contrast to other unicellular fungi, such as Saccharomyces cerevisiae, S. pombe is uniquely devoid of proteins with more than 10 Q repeats. We found that polyQ Htt forms aggregates within S. pombe cells only with exceedingly long polyQ expansions. Surprisingly, despite the presence of polyQ Htt aggregates in both the cytoplasm and nucleus, no significant growth defect was observed in S. pombe cells. Further, PCR analysis showed that the repetitive polyQ-encoding DNA region remained constant following transformation and after multiple divisions in S. pombe, in contrast to the genetic instability of polyQ DNA sequences in other organisms. These results demonstrate that cells with a low content of polyQ or other aggregation-prone proteins can show a striking resilience with respect to polyQ toxicity and that genetic instability of repetitive DNA sequences may have played an important role in the evolutionary emergence and exclusion of polyQ expansion proteins in different organisms.

  12. Mathematical modelling of the corrosion and leaching behaviour of cemented waste forms

    International Nuclear Information System (INIS)

    Kienzler, B.

    1985-05-01

    A theoretical model is presented which allows to calculate the leaching of radionuclides and the corrosion of cemented waste forms in contact with water or brine. The model computes both the behaviour of specimens in laboratory-scale experiments and provides a forecast of the behaviour of waste forms in the case of an accidental drowning of a repository. The mathematical formalism employed describes leaching and corrosion on the basis of diffusion and dissolution processes and of chemical reactions. The mathematical formalism is coded in FORTRAN77. This report includes the documentation of the 'DIFMOD' computer code with the associated 'DIFPLO' plot program and the input manual of both programs. Finally application of the model is demonstrated by some examples allowing interpretation of experimental data. (orig.) [de

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

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    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.

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

    International Nuclear Information System (INIS)

    BLACKFORD LT

    2008-01-01

    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

  15. Development and characterization of new high-level waste form containing LiCl KCl eutectic salts for achieving waste minimization from pyroprocessing

    International Nuclear Information System (INIS)

    Cho, Yong Zun; Kim, In Tae; Park, Hwan Seo; Ahn, Byeung Gil; Eun, Hee Chul; Son, Seock Mo; Ah, Su Na

    2011-12-01

    The purpose of this project is to develop new high level waste (HLW) forms and fabrication processes to dispose of active metal fission products that are removed from electrorefiner salts in the pyroprocessing based fuel cycle. The current technology for disposing of active metal fission products in pyroprocessing involves non selectively discarding of fission product loaded salt in a glass-bonded sodalite ceramic waste form. Selective removal of fission products from the molten salt would greatly minimize the amount of HLW generated and methods were developed to achieve selective separation of fission products during a previous I NERI research project (I NERI 2006 002 K). This I NERI project proceeds from the previous project with the development of suitable waste forms to immobilize the separated fission products. The Korea Atomic Energy Research Institute (KAERI) has focused primarily on developing these waste forms using surrogate waste materials, while the Idaho National Laboratory (INL) has demonstrated fabrication of these waste forms using radioactive electrorefiner salts in hot cell facilities available at INL. Testing and characterization of these radioactive materials was also performed to determine the physical, chemical, and durability properties of the waste forms

  16. Effect of different glass and zeolite A compositions on the leach resistance of ceramic waste forms

    International Nuclear Information System (INIS)

    Lewis, M.A.; Hash, M.; Glandorf, D.

    1996-01-01

    A ceramic waste form is being developed for waste generated during electrometallurgical treatment of spent nuclear fuel. The waste is generated when fission products are removed from the electrolyte, LiCl-KCl eutectic. The waste form is a composite fabricated by hot isostatic pressing a mixture of glass frit and zeolite occluded with fission products and salt. Normalized release rate is less than 1 g/m 2 d for all elements in MCC-1 leach test run for 28 days in deionized water at 90 C. This leach resistance is comparable to that of early Savannah River glasses. We are investigating how leach resistance is affected by changes in cationic form of zeolite and in glass composition. Composites were made with 3 forms of zeolite A and 6 glasses. We used 3-day ASTM C1220-92 (formerly MCC-1) leach tests to screen samples for development purposes only. The leach test results show that the glass composites of zeolites 5A and 4A retain fission products equally well. Loss of Cs is small (0.1-0.5 wt%), while the loss of divalent and trivalent fission products is one or more orders of magnitude smaller. Composites of 5A retain chloride ion better in these short-term screens than 4A and 3A. The more leach resistant composites were made with durable glasses rich in silica and poor in alkaline earth oxides. XRD show that a salt phase was absent in the leach resistant composites of 5A and the better glasses but was present in the other composites with poorer leach performance. Thus, absence of salt phase corresponds to improved leach resistance. Interactions between zeolite and glass depend on composition of both

  17. 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

  18. Preparation, Characterization and Thermal Properties of Paraffin Wax – Expanded Perlite Form-Stable Composites for Latent Heat Storage

    Directory of Open Access Journals (Sweden)

    Tugba GURMEN OZCELIK

    2017-02-01

    Full Text Available In this study, form-stable composite phase change materials (PCM for latent heat storage were prepared by impregnating paraffin wax into the pores of the expanded perlite (EP. The characterization of the composite PCMs was performed by FTIR, TGA, SEM and DSC analysis. The melting point and heat of fusion were determined for 25 % paraffin included composite, as 54.3 °C and 94.71 J/g and for 45 % paraffin included composite as 53.6 °C and 106.69 J/g, respectively. The FTIR results showed that there were no chemical reaction between the perlite and paraffin. TGA analysis indicated that both composite PCMs had good thermal stability. SEM images showed that the paraffin was dispersed uniformly into the pores and on the EP surface. There was no leakage and degradation at the composite PCMs after heating and cooling cycles. According to the results, both prepared composites showed good thermal energy storage properties, reliability and stability. All results suggested that the presented form- stable composite PCMs has great potential for thermal energy storage applications.DOI: http://dx.doi.org/10.5755/j01.ms.23.1.13661

  19. Epoxy-Based Organogels for Thermally Reversible Light Scattering Films and Form-Stable Phase Change Materials.

    Science.gov (United States)

    Puig, Julieta; Dell' Erba, Ignacio E; Schroeder, Walter F; Hoppe, Cristina E; Williams, Roberto J J

    2017-03-29

    Alkyl chains of β-hydroxyesters synthesized by the capping of terminal epoxy groups of diglycidylether of bisphenol A (DGEBA) with palmitic (C16), stearic (C18), or behenic (C22) fatty acids self-assemble forming a crystalline phase. Above a particular concentration solutions of these esters in a variety of solvents led to supramolecular (physical) gels below the crystallization temperature of alkyl chains. A form-stable phase change material (FS-PCM) was obtained by blending the ester derived from behenic acid with eicosane. A blend containing 20 wt % ester was stable as a gel up to 53 °C and exhibited a heat storage capacity of 161 J/g, absorbed during the melting of eicosane at 37 °C. Thermally reversible light scattering (TRLS) films were obtained by visible-light photopolymerization of poly(ethylene glycol) dimethacrylate-ester blends (50 wt %) in the gel state at room temperature. The reaction was very fast and not inhibited by oxygen. TRLS films consisted of a cross-linked methacrylic network interpenetrated by the supramolecular network formed by the esters. Above the melting temperature of crystallites formed by alkyl chains, the film was transparent due to the matching between refractive indices of the methacrylic network and the amorphous ester. Below the crystallization temperature, the film was opaque because of light dispersion produced by the organic crystallites uniformly dispersed in the material. Of high significance for application was the fact that the contrast ratio did not depend on heating and cooling rates.

  20. 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.

  1. 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.

  2. Metal waste forms from treatment of EBR-II spent fuel

    International Nuclear Information System (INIS)

    Abraham, D. P.

    1998-01-01

    Demonstration of Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel is currently being conducted on irradiated, metallic driver fuel and blanket fuel elements from the Experimental Breeder Reactor-II (EBR-II) in Idaho. The residual metallic material from the electrometallurgical treatment process is consolidated into an ingot, the metal waste form (MWF), by employing an induction furnace in a hot cell. Scanning electron microscopy (SEM) and chemical analyses have been performed on irradiated cladding hulls from the driver fuel, and on samples from the alloy ingots. This paper presents the microstructures of the radioactive ingots and compares them with observations on simulated waste forms prepared using non-irradiated material. These simulated waste forms have the baseline composition of stainless steel - 15 wt % zirconium (SS-15Zr). Additions of noble metal elements, which serve as surrogates for fission products, and actinides are made to that baseline composition. The partitioning of noble metal and actinide elements into alloy phases and the role of zirconium for incorporating these elements is discussed in this paper

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

    International Nuclear Information System (INIS)

    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.

    2013-01-01

    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 6 (the Hanford IDF criteria for Na) in the first few hours. The granular and monolithic waste forms also pass the EPA Toxicity Characteristic Leaching Procedure (TCLP) for all Resource Conservation and Recovery Act (RCRA) components at the Universal Treatment

  4. Emergence of interest groups on hazardous waste siting: How do they form and survive

    International Nuclear Information System (INIS)

    Williams, R.G.; Payne, B.A.

    1986-01-01

    The disposal and siting of hazardous and radioactive wastes has created numerous problems for decision-makers in the field of waste management. The social/political problems have proven to be some of the most difficult to solve. Public knowledge of the presence of hazardous and radioactive waste sites has grown considerably in recent years. Over the same period, the process of choosing new disposal sites has attracted a great deal of publicity. In many cases, when existing sites are discovered or when a community is being considered for a new disposal site, organized groups emerge in the community to support or oppose the proposed actions and the decision-makers responsible. Emergent groups are a form of organized collective action in response to a particular situation or event, such as the siting or discovery of a hazardous waste disposal site. Sociological methods and theory can provide insight on the patterns common to these groups, their emergence, and their survival or decline. The questions addressed in this paper are: what are the variables that lead to the formation of such groups, and what conditions or group actions contribute to their growth and survival?

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    J. McClure

    2001-03-12

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

  7. Dechlorination of organochloride waste mixture by microwave irradiation before forming solid recovered fuel.

    Science.gov (United States)

    Liu, Zhen; Wang, Han-Qing; Zhang, Xiao-Dong; Liu, Jian-Wen; Zhou, Yue-Yun

    2017-04-01

    In order to form a modified solid recovered fuel (SRF) with low chlorine content, high calorific value and well combustion performance, low temperature microwave irradiation was applied to remove the chlorine of the organochloride waste mixture before they were mixed to form SRF. The optimizing conditions of final temperature, microwave absorbents and heating rate were also detected to obtain high dechlorination ratio and high ratio of hydrogen chloride (HCl) to volatiles. In the temperature range of 220-300°C, 280°C would be chose as the optimal low microwave modified temperature concerning at which the dechlorination ratio was high and ratio of HCl to volatiles was relatively high as well; The use of microwave absorbents of graphite and silicon carbide (SiC) had a pronounced effect on the dechlorination of organochloride waste mixture, and the dechlorination ratio was increased significantly which could be reached to 87%, almost 20% higher than absorbent absent sample; The heating rate should set be not too fast nor too slow, and there was no big difference between the heating rate of 13°C/min and 15°C/min; The content of Cl of modified SRF is dramatically decreased and reaches to a low level 0.328%. Hence, the modified SRF can be ascended from the third class to the second class according to the Finland chlorine Classes I-III. Moreover, the combustibility of modified SRF was substantial improved compared to the traditional SRF. The low heating value was almost 20.56MJ/kg which is close to the LHV of lignite coal and bituminous coal in China, and it increased by 60% over that of traditional SRF. Removing chlorine of organochloride waste mixture before they are mixed with other kinds of combustible waste to form a modified SRF which is expected to be an alternative fuel for combustion in the future. Copyright © 2016 Elsevier Ltd. All rights reserved.

  8. 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

  9. Development of form stable Poly(methyl methacrylate) (PMMA) coated thermal phase change material for solar water heater applications

    Science.gov (United States)

    Munusamy, Y.; Shanmugam, S.; Shi-Ying, Kee

    2018-04-01

    Phase change material (PCM) is one of the most popular and widely used thermal energy storage material in solar water heater because it able to absorb and release a large amount of latent heat during a phase change process over a narrow temperature range. However the practical application of PCM is limited by two major issues; 1) leakage which leads to material loss and corrosion of tank and 2) large volume change during phase change process which cause pressure build up in the tank. In this work, form-stable PCM was prepared by coating myristic acid with Poly(methyl methacrylate) (PMMA) to prevent leakage of PCM. PMMA was mixed with different weight percentage (0.1, 0.2, 0.3, 0.4 and 0.5 wt%) of dicumyl peroxide (DCP). The purpose of adding DCP to PMMA is to crosslink the polymer and to increase the mechanical strength of PMMA to hold the myristic acid content inside the coating during the phase change process. Leakage test results showed that PMMA mixed with 0.1% DCP exhibit 0% leakage. This result is further supported by Field Emission Scanning Electron Microscopy (FESEM) images and Fourier transform infrared spectroscopy (FTIR) analysis results, where a compact and uniform coating without cracks were formed for PCM coated with PMMA with 0.1% DCP. Differential scanning calorimetry (DSC) results shows that the melting point of form-stable PCM is 55°C, freezing point is 50°C, the latent heat of melting and freezing is 67.59 J/g.

  10. Method for making a low density polyethylene waste form for safe disposal of low level radioactive material

    Science.gov (United States)

    Colombo, P.; Kalb, P.D.

    1984-06-05

    In the method of the invention low density polyethylene pellets are mixed in a predetermined ratio with radioactive particulate material, then the mixture is fed through a screw-type extruder that melts the low density polyethylene under a predetermined pressure and temperature to form a homogeneous matrix that is extruded and separated into solid monolithic waste forms. The solid waste forms are adapted to be safely handled, stored for a short time, and safely disposed of in approved depositories.

  11. A Science-Based Approach to Understanding Waste Form Durability in Open and Closed Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

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

    2007-01-01

    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 U 6+ - 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 behavior of the source term over long time periods (greater than 10 5 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. (authors)

  12. A Science-Based Approach to Understanding Waste Form Durability in Open and Closed Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

    M.T. Peters; R.C. Ewing

    2006-01-01

    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 U 6+ -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 behavior of the source term over long time periods (greater than 10 5 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

  13. 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.

  14. Perception and the strongest sensory memory trace of multi-stable displays both form shortly after the stimulus onset.

    Science.gov (United States)

    Pastukhov, Alexander

    2016-02-01

    We investigated the relation between perception and sensory memory of multi-stable structure-from-motion displays. The latter is an implicit visual memory that reflects a recent history of perceptual dominance and influences only the initial perception of multi-stable displa