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Sample records for cementitious waste form

  1. 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.

  2. Heat of Hydration of Low Activity Cementitious Waste Forms

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

    2015-07-23

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

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

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    Harbour, J

    2007-03-02

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Harbour, J

    2007-03-02

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

  5. DEMONSTRATION OF LEACHXS/ORCHESTRA CAPABILITIES BY SIMULATING CONSTITUENT RELEASE FROM A CEMENTITIOUS WASTE FORM IN A REINFORCED CONCRETE VAULT

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    Langton, C.; Meeussen, J.; Sloot, H.

    2010-03-31

    The objective of the work described in this report is to demonstrate the capabilities of the current version of LeachXS{trademark}/ORCHESTRA for simulating chemical behavior and constituent release processes in a range of applications that are relevant to the CBP. This report illustrates the use of LeachXS{trademark}/ORCHESTRA for the following applications: (1) Comparing model and experimental results for leaching tests for a range of cementitious materials including cement mortars, grout, stabilized waste, and concrete. The leaching test data includes liquid-solid partitioning as a function of pH and release rates based on laboratory column, monolith, and field testing. (2) Modeling chemical speciation of constituents in cementitious materials, including liquid-solid partitioning and release rates. (3) Evaluating uncertainty in model predictions based on uncertainty in underlying composition, thermodynamic, and transport characteristics. (4) Generating predominance diagrams to evaluate predicted chemical changes as a result of material aging using the example of exposure to atmospheric conditions. (5) Modeling coupled geochemical speciation and diffusion in a three layer system consisting of a layer of Saltstone, a concrete barrier, and a layer of soil in contact with air. The simulations show developing concentration fronts over a time period of 1000 years. (6) Modeling sulfate attack and cracking due to ettringite formation. A detailed example for this case is provided in a separate article by the authors (Sarkar et al. 2010). Finally, based on the computed results, the sensitive input parameters for this type of modeling are identified and discussed. The chemical speciation behavior of substances is calculated for a batch system and also in combination with transport and within a three layer system. This includes release from a barrier to the surrounding soil as a function of time. As input for the simulations, the physical and chemical properties of the

  6. Cementitious Stabilization of Mixed Wastes with High Salt Loadings

    Energy Technology Data Exchange (ETDEWEB)

    Spence, R.D.; Burgess, M.W.; Fedorov, V.V.; Downing, D.J.

    1999-04-01

    Salt loadings approaching 50 wt % were tolerated in cementitious waste forms that still met leach and strength criteria, addressing a Technology Deficiency of low salt loadings previously identified by the Mixed Waste Focus Area. A statistical design quantified the effect of different stabilizing ingredients and salt loading on performance at lower loadings, allowing selection of the more effective ingredients for studying the higher salt loadings. In general, the final waste form needed to consist of 25 wt % of the dry stabilizing ingredients to meet the criteria used and 25 wt % water to form a workable paste, leaving 50 wt % for waste solids. The salt loading depends on the salt content of the waste solids but could be as high as 50 wt % if all the waste solids are salt.

  7. Cementitious waste option scoping study report

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    Lee, A.E.; Taylor, D.D.

    1998-02-01

    A Settlement Agreement between the Department of Energy (DOE) and the State of Idaho mandates that all high-level radioactive waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) on the Idaho National Engineering and Environmental Laboratory (INEEL) will be treated so that it is ready to be moved out of Idaho for disposal by a target date of 2035. This study investigates the nonseparations Cementitious Waste Option (CWO) as a means to achieve this goal. Under this option all liquid sodium-bearing waste (SBW) and existing HLW calcine would be recalcined with sucrose, grouted, canisterized, and interim stored as a mixed-HLW for eventual preparation and shipment off-Site for disposal. The CWO waste would be transported to a Greater Confinement Disposal Facility (GCDF) located in the southwestern desert of the US on the Nevada Test Site (NTS). All transport preparation, shipment, and disposal facility activities are beyond the scope of this study. CWO waste processing, packaging, and interim storage would occur over a 5-year period between 2013 and 2017. Waste transport and disposal would occur during the same time period.

  8. Obtaining cementitious material from municipal solid waste

    Directory of Open Access Journals (Sweden)

    Macías, A.

    2007-06-01

    Full Text Available The primary purpose of the present study was to determine the viability of using incinerator ash and slag from municipal solid waste as a secondary source of cementitious materials. The combustion products used were taken from two types of Spanish MSW incinerators, one located at Valdemingómez, in Madrid, and the other in Melilla, with different incineration systems: one with fluidised bed combustion and other with mass burn waterwall. The effect of temperature (from 800 to 1,200 ºC on washed and unwashed incinerator residue was studied, in particular with regard to phase formation in washed products with a high NaCl and KCl content. The solid phases obtained were characterized by X-ray diffraction and BET-N2 specific surface procedures.El principal objetivo del trabajo ha sido determinar la viabilidad del uso de las cenizas y escorias procedentes de la incineración de residuos sólidos urbanos, como materia prima secundaria para la obtención de fases cementantes. Para ello se han empleado los residuos generados en dos tipos de incineradoras españolas de residuos sólidos urbanos: la incineradora de Valdemingómez y la incineradora de Melilla. Se ha estudiado la transformación de los residuos, sin tratamiento previo, en función de la temperatura de calentamiento (desde 800 ºC hasta 1.200 ºC, así como la influencia del lavado de los residuos con alto contenido en NaCl y KCl en la formación de fases obtenidas a las diferentes temperaturas de calcinación. Las fases obtenidas fueron caracterizadas por difracción de rayos X y área superficial por el método BET-N2.

  9. Direct cementitious waste option study report

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    Dafoe, R.E.; Losinski, S.J.

    1998-02-01

    A settlement agreement between the Department of Energy (DOE) and the State of Idaho mandates that all high-level radioactive waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) will be treated so that it is ready to be moved out of Idaho for disposal by a target data of 2035. This study investigates the direct grouting of all ICPP calcine (including the HLW dry calcine and those resulting from calcining sodium-bearing liquid waste currently residing in the ICPP storage tanks) as the treatment method to comply with the settlement agreement. This method involves grouting the calcined waste and casting the resulting hydroceramic grout into stainless steel canisters. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory (INEEL) until they are sent to a national geologic repository. The operating period for grouting treatment will be from 2013 through 2032, and all the HLW will be treated and in interim storage by the end of 2032.

  10. Investigations on cementitious composites based on rubber particle waste additions

    Directory of Open Access Journals (Sweden)

    Glaucio Laun Nacif

    2013-04-01

    Full Text Available The amount of waste rubber has gradually increased over recent years because of over-growing use of rubber products. The disposal of waste rubber has caused serious environmental problems. The incorporation of recycled materials into cementitious composites is a feasible alternative that has gained ground in civil construction. The performance of such materials is much affected not only by the rubber addition, but also the particle size which has been controversially reported in the literature. In order to investigate the single effect of rubber particles into cement based materials, rubber cementitious composites were prepared with no silica particle additions. A full factorial design has been conducted to assess the influence of the rubber particle size (0.84/0.58 mm and 0.28/0.18 mm; mass fraction used (5, 15 and 30%; and water/cement ratio (0.35 and 0.50 on the physic-mechanical properties of the composites. The materials were characterized through apparent density, porosity, compressive strength, flexural strength, modulus of elasticity and microstructural analysis. The interactions of rubber particle size, rubber fraction and water/cement ratio affected significantly the density and compressive strength of the composites. The apparent porosity was influenced mainly by the rubber particle size. The flexural strength was affected by the main factors and the modulus of elasticity was affected by the interaction factors rubber particle size and fraction, and rubber fraction and w/c ratio.

  11. CEMENTITIOUS GROUT FOR CLOSING SRS HIGH LEVEL WASTE TANKS - #12315

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    Langton, C.; Burns, H.; Stefanko, D.

    2012-01-10

    In 1997, the first two United States Department of Energy (US DOE) high level waste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-level waste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. The closure will also fill, physically stabilize and isolate ancillary equipment abandoned in the tanks. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and chemically reduction potential (Eh) of -200 to -400 to stabilize selected potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted, respectively, to support the mass placement strategy developed by

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

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    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.

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

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

    1995-07-01

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

  14. 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

  15. The Cementitious Barriers Partnership Experimental Programs and Software Advancing DOE’s Waste Disposal/Tank Closure Efforts – 15436

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    Burns, Heather [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Flach, Greg [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith, Frank [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Langton, Christine [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Brown, Kevin [Vanderbilt Univ./CRESP, Nashville, TN (United States); Kosson, David [Vanderbilt Univ./CRESP, Nashville, TN (United States); Samson, Eric [SIMCO Technologies, Inc. (United States); Mallick, Pramod [US DOE, Washington, DC (United States)

    2015-01-27

    The U.S. Department of Energy Environmental Management (DOE-EM) Office of Tank Waste Management-sponsored Cementitious Barriers Partnership (CBP) is chartered with providing the technical basis for implementing cement-based waste forms and radioactive waste containment structures for long-term disposal. DOE needs in this area include the following to support progress in final treatment and disposal of legacy waste and closure of High-Level Waste (HLW) tanks in the DOE complex: long-term performance predictions, flow sheet development and flow sheet enhancements, and conceptual designs for new disposal facilities. The DOE-EM Cementitious Barriers Partnership is producing software and experimental programs resulting in new methods and data needed for end-users involved with environmental cleanup and waste disposal. Both the modeling tools and the experimental data have already benefited the DOE sites in the areas of performance assessments by increasing confidence backed up with modeling support, leaching methods, and transport properties developed for actual DOE materials. In 2014, the CBP Partnership released the CBP Software Toolbox –“Version 2.0” which provides concrete degradation models for 1) sulfate attack, 2) carbonation, and 3) chloride initiated rebar corrosion, and includes constituent leaching. These models are applicable and can be used by both DOE and the Nuclear Regulatory Commission (NRC) for service life and long-term performance evaluations and predictions of nuclear and radioactive waste containment structures across the DOE complex, including future SRS Saltstone and HLW tank performance assessments and special analyses, Hanford site HLW tank closure projects and other projects in which cementitious barriers are required, the Advanced Simulation Capability for Environmental Management (ASCEM) project which requires source terms from cementitious containment structures as input to their flow simulations, regulatory reviews of DOE performance

  16. 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.

  17. Reuse of Woody Biomass Ash Waste in Cementitious Materials

    OpenAIRE

    Ukrainczyk, N.; Vrbos, N.; Koenders, E.A.B.

    2016-01-01

    There is an increased interest in the reuse of ash waste from biomass combustion, being a sustainable source of energy. This paper investigates the partial replacement of cement and sand in building materials with fly ash waste generated from combustion of woody biomass waste. The results show that the ash widens the particle size distribution of cement and has minerals complementary to portland cement, thus justifying its application as cement replacement, but with a relatively high amoun...

  18. Densified waste form and method for forming

    Science.gov (United States)

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

    2015-08-25

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

  19. A new alkali-activated steel slag-based cementitious material for photocatalytic degradation of organic pollutant from waste water.

    Science.gov (United States)

    Zhang, Yao Jun; Liu, Li Cai; Xu, Yong; Wang, Ya Chao; Xu, De Long

    2012-03-30

    A new type of Ni,Ca-cementitious material was firstly synthesized via a two-step reaction of alkali-activated steel slag polymerization and ion exchange. The XRF results showed that almost all the Na(+) ions in the matrix of Na,Ca-cementitious material were replaced by Ni(2+) ions at room temperature. The new hydrated products of metahalloysite (Si(2)Al(2)O(5)(OH)(4)) and calcium silicate hydrate (CSH) were formed in the Na,Ca-cementitious material. The diffuse reflectance UV-vis near infrared ray spectrum was blue-shifted due to the strong interaction between Ni(2+) and negative charge of [AlO(4)](5-) tetrahedron in the framework of cementitious material. The Ni,Ca-cementitious material was used as a catalyst for the photocatalytic degradation of methylene blue dye and showed a degradation rate of 94.39% under UV irradiation. The high photocatalytic degradation activity was suggested to be the synergistic effect of the cementitious matrix, Ni(2+) ions and the iron oxides of wustite (FeO) and calcium iron oxide (Ca(2)Fe(2)O(5)) from the steel slag. A probable mechanism of photocatalytic oxidative degradation was proposed.

  20. The solubility of nickel and its migration through the cementitious backfill of a geological disposal facility for nuclear waste.

    Science.gov (United States)

    Felipe-Sotelo, M; Hinchliff, J; Field, L P; Milodowski, A E; Holt, J D; Taylor, S E; Read, D

    2016-08-15

    This work describes the solubility of nickel under the alkaline conditions anticipated in the near field of a cementitious repository for intermediate level nuclear waste. The measured solubility of Ni in 95%-saturated Ca(OH)2 solution is similar to values obtained in water equilibrated with a bespoke cementitious backfill material, on the order of 5×10(-7)M. Solubility in 0.02M NaOH is one order of magnitude lower. For all solutions, the solubility limiting phase is Ni(OH)2; powder X-ray diffraction and scanning transmission electron microscopy indicate that differences in crystallinity are the likely cause of the lower solubility observed in NaOH. The presence of cellulose degradation products causes an increase in the solubility of Ni by approximately one order of magnitude. The organic compounds significantly increase the rate of Ni transport under advective conditions and show measurable diffusive transport through intact monoliths of the cementitious backfill material.

  1. A new alkali-activated steel slag-based cementitious material for photocatalytic degradation of organic pollutant from waste water

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yao Jun, E-mail: yaojzhang@yahoo.com.cn [College of Material Science and Engineering, Xi' an University of Architecture and Technology, Xi' an 710055 (China); Liu, Li Cai; Xu, Yong; Wang, Ya Chao; Xu, De Long [College of Material Science and Engineering, Xi' an University of Architecture and Technology, Xi' an 710055 (China)

    2012-03-30

    Highlights: Black-Right-Pointing-Pointer A novel Ni,Ca-cementitious material is synthesized by a two-step reaction. Black-Right-Pointing-Pointer Ni,Ca-geopolymer is firstly used for the photocatalytic degradation of MB. Black-Right-Pointing-Pointer Absorption bands in the UV and NIR regions are reported for the first time. Black-Right-Pointing-Pointer A reaction mechanism of photocatalytic degradation was proposed. - Abstract: A new type of Ni,Ca-cementitious material was firstly synthesized via a two-step reaction of alkali-activated steel slag polymerization and ion exchange. The XRF results showed that almost all the Na{sup +} ions in the matrix of Na,Ca-cementitious material were replaced by Ni{sup 2+} ions at room temperature. The new hydrated products of metahalloysite (Si{sub 2}Al{sub 2}O{sub 5}(OH){sub 4}) and calcium silicate hydrate (CSH) were formed in the Na,Ca-cementitious material. The diffuse reflectance UV-vis near infrared ray spectrum was blue-shifted due to the strong interaction between Ni{sup 2+} and negative charge of [AlO{sub 4}]{sup 5-} tetrahedron in the framework of cementitious material. The Ni,Ca-cementitious material was used as a catalyst for the photocatalytic degradation of methylene blue dye and showed a degradation rate of 94.39% under UV irradiation. The high photocatalytic degradation activity was suggested to be the synergistic effect of the cementitious matrix, Ni{sup 2+} ions and the iron oxides of wustite (FeO) and calcium iron oxide (Ca{sub 2}Fe{sub 2}O{sub 5}) from the steel slag. A probable mechanism of photocatalytic oxidative degradation was proposed.

  2. Utilization of Construction Waste Composite Powder Materials as Cementitious Materials in Small-Scale Prefabricated Concrete

    Directory of Open Access Journals (Sweden)

    Cuizhen Xue

    2016-01-01

    Full Text Available The construction and demolition wastes have increased rapidly due to the prosperity of infrastructure construction. For the sake of effectively reusing construction wastes, this paper studied the potential use of construction waste composite powder material (CWCPM as cementitious materials in small-scale prefabricated concretes. Three types of such concretes, namely, C20, C25, and C30, were selected to investigate the influences of CWCPM on their working performances, mechanical properties, and antipermeability and antifrost performances. Also the effects of CWCPM on the morphology, hydration products, and pore structure characteristics of the cement-based materials were analyzed. The results are encouraging. Although CWCPM slightly decreases the mechanical properties of the C20 concrete and the 7 d compressive strengths of the C25 and C30 concretes, the 28 d compressive strength and the 90 d flexural strength of the C25 and C30 concretes are improved when CWCPM has a dosage less than 30%; CWCPM improves the antipermeability and antifrost performances of the concretes due to its filling and pozzolanic effects; the best improvement is obtained at CWCPM dosage of 30%; CWCPM optimizes cement hydration products, refines concrete pore structure, and gives rise to reasonable pore size distribution, therefore significantly improving the durability of the concretes.

  3. DURABILITY OF GREEN CONCRETE WITH TERNARY CEMENTITIOUS SYSTEM CONTAINING RECYCLED AGGREGATE CONCRETE AND TIRE RUBBER WASTES

    Directory of Open Access Journals (Sweden)

    MAJID MATOUQ ASSAS

    2016-06-01

    Full Text Available All over the world billions of tires are being discarded and buried representing a serious ecological threat. Up to now a small part is recycled and millions of tires are just stockpiled, landfilled or buried. This paper presents results about the properties and the durability of green concrete contains recycled concrete as a coarse aggregate with partial replacement of sand by tire rubber wastes for pavement use. Ternary cementious system, Silica fume, Fly ash and Cement Kiln Dust are used as partial replacement of cement by weight. Each one replaced 10% of cement weight to give a total replacement of 30%. The durability performance was assessed by means of water absorption, chloride ion permeability at 28 and 90 days, and resistance to sulphuric acid attack at 1, 7, 14 and 28 days. Also to the compression behaviors for the tested specimens at 7, 14, 28 and 90 days were detected. The results show the existence of ternary cementitious system, silica fly ash and Cement Kiln Dust minimizes the strength loss associated to the use of rubber waste. In this way, up to 10% rubber content and 30% ternary cementious system an adequate strength class value (30 MPa, as required for a wide range of common structural uses, can be reached both through natural aggregate concrete and recycled aggregate concrete. Results also show that, it is possible to use rubber waste up to 15% and still maintain a high resistance to acid attack. The mixes with 10%silica fume, 10% fly ash and 10% Cement Kiln Dust show a higher resistance to sulphuric acid attack than the reference mix independently of the rubber waste content. The mixes with rubber waste and ternary cementious system was a lower resistance to sulphuric acid attack than the reference mix.

  4. Comparative waste forms study

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-12-01

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

  5. 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.

  6. Long-term degradation (or improvement?) of cementitious grout/concrete for waste disposal at Hanford

    Energy Technology Data Exchange (ETDEWEB)

    Piepho, M.G. [Daniel B. Stephens & Associates, Inc., Richland, WA (United States)

    1997-12-31

    If grout and/or concrete barriers and containments are considered for long-term (500 yrs to 100,000 ) waste disposal, then long-term degradation of grout/cement materials (and others) need to be studied. Long-term degradations of a cementitious grout monolith (15.4mW x 10.4mH x 37.6mL) and its containment concrete shell and asphalt shell (each 1-m thick) were analyzed. The main degradation process of the concrete shell was believed to be fractures due to construction joints, shrinkage, thermal stress, settlement, and seismic events. A scenario with fractures was modeled (flow and transport model) for long-term risk performance (out to a million yrs). Even though the concrete/grout is expected to fracture, the concrete/grout chemistry, which has high Ph value, is very beneficial in causing calcite deposits from calcium in the water precipitating in the fractures. These calcite deposits will tend to plug the fracture and keep water from entering. The effectiveness of such plugging needs to be studied more. It`s possible that the plugged fractures are more impermeable than the original concrete/grout. The long-term performance of concrete/grout barriers will be determined by its chemistry, not its mechanical properties.

  7. Photovoltaic's silica-rich waste sludge as supplementary cementitious material (SCM)

    Energy Technology Data Exchange (ETDEWEB)

    Quercia, G., E-mail: g.quercia@tue.nl [Materials innovation institute (M2i), Mekelweg 2, P.O. Box 5008, 2600 GA Delft (Netherlands); Eindhoven University of Technology, Department of the Built Environment, P.O. Box 513, 5600 MB Eindhoven (Netherlands); Putten, J.J.G. van der [Eindhoven University of Technology, Department of the Built Environment, P.O. Box 513, 5600 MB Eindhoven (Netherlands); Hüsken, G. [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 87, D-12205 Berlin (Germany); Brouwers, H.J.H. [Eindhoven University of Technology, Department of the Built Environment, P.O. Box 513, 5600 MB Eindhoven (Netherlands)

    2013-12-15

    Waste sludge, a solid recovered from wastewater of photovoltaic-industries, composes of agglomerates of nano-particles like SiO{sub 2} and CaCO{sub 3}. This sludge deflocculates in aqueous solutions into nano-particles smaller than 1 μm. Thus, this sludge constitutes a potentially hazardous waste when it is improperly disposed. Due to its high content of amorphous SiO{sub 2}, this sludge has a potential use as supplementary cementitious material (SCM) in concrete. In this study the main properties of three different samples of photovoltaic's silica-rich waste sludge (nSS) were physically and chemically characterized. The characterization techniques included: scanning electron microscopy (SEM), X-ray energy dispersive spectroscopy (EDS), X-ray diffraction (XRD), nitrogen physical adsorption isotherm (BET method), density by Helium pycnometry, particle size distribution determined by laser light scattering (LLS) and zeta-potential measurements by dynamic light scattering (DLS). In addition, a dispersability study was performed to design stable slurries to be used as liquid additives for the concrete production on site. The effects on the hydration kinetics of cement pastes by the incorporation of nSS in the designed slurries were determined using an isothermal calorimeter. A compressive strength test of standard mortars with 7% of cement replacement was performed to determine the pozzolanic activity of the waste nano-silica sludge. Finally, the hardened system was fully characterized to determine the phase composition. The results demonstrate that the nSS can be utilized as SCM to replace portion of cement in mortars, thereby decreasing the CO{sub 2} footprint and the environmental impact of concrete. -- Highlights: •Three different samples of PV nano-silica sludge (nSS) were fully characterized. •nSS is composed of agglomerates of nano-particles like SiO{sub 2} and CaCO{sub 3}. •Dispersability studies demonstrated that nSS agglomerates are broken to nano

  8. 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.

  9. Self-healing behavior of strain hardening cementitious composites incorporating local waste materials

    NARCIS (Netherlands)

    Qian, S.; Zhou, J.; Rooij, M.R. de; Schlangen, E.; Ye, G.; Breugel, K. van

    2009-01-01

    The self-healing behavior of a series of pre-cracked fiber reinforced strain hardening cementitious composites incorporating blast furnace slag (BFS) and limestone powder (LP) with relatively high water/binder ratio is investigated in this paper, focusing on the recovery of its deflection capacity.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-30

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

  11. Intended long term performances of cementitious engineered barriers for future storage and disposal facilities for radioactive wastes in Romania

    Directory of Open Access Journals (Sweden)

    Sociu F.

    2013-07-01

    Full Text Available Considering the EU statements, Romania is engaged to endorse in the near future the IAEA relevant publications on geological repository (CNCANa, to update the Medium and Long Term National Strategy for Safe Management of Radioactive Waste and to approve the Road Map for Geological Repository Development. Currently, for example, spent fuel is wet stored for 6 years and after this period it is transported to dry storage in MACSTOR-200 (a concrete monolithic module where it is intended to remain at least 50 years. The present situation for radioactive waste management in Romania is reviewed in the present paper. Focus will be done on existent disposal facilities but, also, on future facilities planned for storage / disposal of radioactive wastes. Considering specific data for Romanian radioactive waste inventory, authors are reviewing the advance in the radioactive waste management in Romania considering its particularities. The team tries to highlight the expected limitations and unknown data related with cementitious engineered barriers that has to be faced in the near future incase of interim storage or for the upcoming long periods of disposal.

  12. Intended long term performances of cementitious engineered barriers for future storage and disposal facilities for radioactive wastes in Romania

    Science.gov (United States)

    Fako, R.; Barariu, Gh.; Toma, R.; Georgescu, R.; Sociu, F.

    2013-07-01

    Considering the EU statements, Romania is engaged to endorse in the near future the IAEA relevant publications on geological repository (CNCANa), to update the Medium and Long Term National Strategy for Safe Management of Radioactive Waste and to approve the Road Map for Geological Repository Development. Currently, for example, spent fuel is wet stored for 6 years and after this period it is transported to dry storage in MACSTOR-200 (a concrete monolithic module) where it is intended to remain at least 50 years. The present situation for radioactive waste management in Romania is reviewed in the present paper. Focus will be done on existent disposal facilities but, also, on future facilities planned for storage / disposal of radioactive wastes. Considering specific data for Romanian radioactive waste inventory, authors are reviewing the advance in the radioactive waste management in Romania considering its particularities. The team tries to highlight the expected limitations and unknown data related with cementitious engineered barriers that has to be faced in the near future incase of interim storage or for the upcoming long periods of disposal.

  13. Guidelines for assessing the valorization of a waste into cementitious material: dredged sediment for production of self compacting concrete

    Directory of Open Access Journals (Sweden)

    Rozas, F.

    2015-09-01

    Full Text Available This article presents some guidelines in order to analyse the feasibility of including a waste material in the production of a structural cementitious material. First of all, the compatibility of the waste with a cementitious material has to be assured; then, if necessary, a decontamination step will be carried out; after, decision on the type of material has to be taken based on different aspects, with special emphasis on the granulometry. As a last step, mechanical, environmental and durability properties have to be evaluated. Then the procedure is illustrated with a full example, obtaining a self compacting concrete (SCC including dredged sediment taken from a Spanish harbour.Este artículo presenta algunas directrices con el fin de analizar la posibilidad de incluir un material de desecho en la producción de un material base cemento estructural. En primer lugar, debe asegurarse la compatibilidad de los residuos con el material base cemento. Tras ello, si es necesario, se llevará a cabo la etapa de descontaminación del residuo. Después debe tomarse la decisión sobre el tipo de material a utilizar en base a diferentes aspectos, haciendo especial énfasis en la granulometría. Como último paso, deben evaluarse las propiedades mecánicas, ambientales y de durabilidad del producto final. El procedimiento a seguir se ilustra con un ejemplo concreto basado en la obtención de un hormigón autocompactante (SCC incluyendo en su fabricación sedimentos dragados tomados de un puerto español.

  14. REFERENCE CASES FOR USE IN THE CEMENTITIOUS BARRIERS PARTNERSHIP

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C

    2009-01-06

    The Cementitious Barriers Project (CBP) is a multidisciplinary cross cutting project initiated by the US Department of Energy (DOE) to develop a reasonable and credible set of tools to improve understanding and prediction of the structural, hydraulic and chemical performance of cementitious barriers used in nuclear applications. The period of performance is >100 years for operating facilities and > 1000 years for waste management. The CBP has defined a set of reference cases to provide the following functions: (1) a common set of system configurations to illustrate the methods and tools developed by the CBP, (2) a common basis for evaluating methodology for uncertainty characterization, (3) a common set of cases to develop a complete set of parameter and changes in parameters as a function of time and changing conditions, and (4) a basis for experiments and model validation, and (5) a basis for improving conceptual models and reducing model uncertainties. These reference cases include the following two reference disposal units and a reference storage unit: (1) a cementitious low activity waste form in a reinforced concrete disposal vault, (2) a concrete vault containing a steel high-level waste tank filled with grout (closed high-level waste tank), and (3) a spent nuclear fuel basin during operation. Each case provides a different set of desired performance characteristics and interfaces between materials and with the environment. Examples of concretes, grout fills and a cementitious waste form are identified for the relevant reference case configurations.

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-15

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

  17. Cementitious Mixtures for Sealing Evaporite and Clastic Rocks in a Radioactive-Waste Repository.

    Science.gov (United States)

    1985-09-01

    on other materials (Struble, Skalny, and Mindess 1980; Barnes, Diamond, and Dolch 1978). Images of chloride distribution show concentrations of...Waste Management, US Department of Energy, Washington, DC. Struble, L., Skalny, L., and Mindess , S. 1980. Cement and Concrete Research, Vol 10, pp

  18. Study on utilization of industrial waste residue for production of microporous cementitious materials%利用工业废渣生产多微孔胶凝材料的研究

    Institute of Scientific and Technical Information of China (English)

    周君生; 徐伟; 陈益兰; 潘荣伟

    2012-01-01

    A new inorganic cementitious material with microporous structure was prepared by single low-temperature sintering process with fly ash,carbide slag,and desulfurization gypsum as raw materials and with adding a small amount of mineralizer,adhesive agent,and pore-forming agent and the material's performances were also studied.Taking the optimized formula and under the conditions of pre-heat temperature 400 ℃, roasting temperature 1 220 ℃, and holding time 25 min, water absorption, apparent density, and 3 d compressive strength of the prepared microporous cementitious material were 2.10% ,1.32 g/cm3,and 8.56 MPa,respectively.XRD analysis showed the main phases of synthesized material were calcium silicate and gehlenite.Using industrial waste slag as raw material as well as the combination of porosity and cementitious character made the microporous cementitious material can partially replace cement and lightweight aggregates (haydite) to prepare insulating mortar so that the purpose of environmental protection and energy saving could be realized.%以粉煤灰、电石渣和脱硫石膏为主要原料,掺加少量矿化剂、黏结剂和造孔剂,采用一次低温烧成工艺,制备出具有多微孔结构的新型无机胶凝材料,并对产品性能进行了研究.采用优化配方,在预热温度为400℃、焙烧温度为1 220℃、保温时间为25 min条件下制备的多微孔胶凝材料,其吸水率为2.10%、表观密度为1.32 g/cm3,3d抗压强度为8.56 MPa.通过XRD分析可知,合成材料的主要矿物相为硅酸钙和钙铝黄长石.材料采用工业废渣制备,集多孔性和胶凝性,可部分代替水泥和陶粒制成保温砂浆,达到环保节能的目的.

  19. Review of the potential effects of alkaline plume migration from a cementitious repository for radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Savage, D.

    1997-09-01

    Extensive use of cement and concrete is envisaged in the construction of geological repositories for low and intermediate-level radioactive wastes, both for structural, and encapsulation and backfilling purposes. Saturation of these materials with groundwater may occur in the post-closure period of disposal, producing a hyperalkaline pore fluid with a pH in the range 10-13.5. These pore fluids have the potential to migrate from the repository according to local groundwater flow conditions and react chemically with the host rock. These chemical reactions may affect the rock`s capacity to retard the migration of radionuclides released from the repository after the degradation of the waste packages. The effects of these chemical reactions on the behaviour of the repository rock as a barrier to waste migration need to be investigated for the purposes of assessing the safety of the repository design (so-called `safety assessment` or `performance assessment`). The objectives of the work reported here were to: identify those processes influencing radionuclide mobility in the geosphere which could be affected by plume migration; review literature relevant to alkali-rock reaction; contact organisations carrying out relevant research and summarise their current and future activities; and make recommendations how the effects of plume migration can be incorporated into models of repository performance assessment. (author).

  20. Development and Demonstration of Material Properties Database and Software for the Simulation of Flow Properties in Cementitious Materials

    Energy Technology Data Exchange (ETDEWEB)

    Smith, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Flach, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-03-30

    This report describes work performed by the Savannah River National Laboratory (SRNL) in fiscal year 2014 to develop a new Cementitious Barriers Project (CBP) software module designated as FLOExcel. FLOExcel incorporates a uniform database to capture material characterization data and a GoldSim model to define flow properties for both intact and fractured cementitious materials and estimate Darcy velocity based on specified hydraulic head gradient and matric tension. The software module includes hydraulic parameters for intact cementitious and granular materials in the database and a standalone GoldSim framework to manipulate the data. The database will be updated with new data as it comes available. The software module will later be integrated into the next release of the CBP Toolbox, Version 3.0. This report documents the development efforts for this software module. The FY14 activities described in this report focused on the following two items that form the FLOExcel package; 1) Development of a uniform database to capture CBP data for cementitious materials. In particular, the inclusion and use of hydraulic properties of the materials are emphasized; and 2) Development of algorithms and a GoldSim User Interface to calculate hydraulic flow properties of degraded and fractured cementitious materials. Hydraulic properties are required in a simulation of flow through cementitious materials such as Saltstone, waste tank fill grout, and concrete barriers. At SRNL these simulations have been performed using the PORFLOW code as part of Performance Assessments for salt waste disposal and waste tank closure.

  1. OVERVIEW OF THE U.S. DEPARTMENT OF ENERGY AND NUCLEAR REGULATORY COMMISSION PERFORMANCE ASSESSMENT APPROACHES: CEMENTITIOUS BARRIERS PARTNERSHIP

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Burns, H.

    2009-05-29

    Engineered barriers including cementitious barriers are used at sites disposing or contaminated with low-level radioactive waste to enhance performance of the natural environment with respect to controlling the potential spread of contaminants. Drivers for using cementitious barriers include: high radionuclide inventory, radionuclide characteristics (e.g., long half-live, high mobility due to chemical form/speciation, waste matrix properties, shallow water table, and humid climate that provides water for leaching the waste). This document comprises the first in a series of reports being prepared for the Cementitious Barriers Partnership. The document is divided into two parts which provide a summary of: (1) existing experience in the assessment of performance of cementitious materials used for radioactive waste management and disposal and (2) sensitivity and uncertainty analysis approaches that have been applied for assessments. Each chapter is organized into five parts: Introduction, Regulatory Considerations, Specific Examples, Summary of Modeling Approaches and Conclusions and Needs. The objective of the report is to provide perspective on the state of the practice for conducting assessments for facilities involving cementitious barriers and to identify opportunities for improvements to the existing approaches. Examples are provided in two contexts: (1) performance assessments conducted for waste disposal facilities and (2) performance assessment-like analyses (e.g., risk assessments) conducted under other regulatory regimes. The introductory sections of each section provide a perspective on the purpose of performance assessments and different roles of cementitious materials for radioactive waste management. Significant experience with assessments of cementitious materials associated with radioactive waste disposal concepts exists in the US Department of Energy Complex and the commercial nuclear sector. Recently, the desire to close legacy facilities has created

  2. Colloids in the mortar backfill of a cementitious repository for radioactive waste.

    Science.gov (United States)

    Wieland, E; Spieler, P

    2001-01-01

    Colloids are present in groundwater aquifers and water-permeable engineered barrier systems and may facilitate the migration of radionuclides. A highly permeable mortar is foreseen to be used as backfill for the engineered barrier of the Swiss repository for low- and intermediate-level waste. The backfill is considered to be a chemical environment with some potential for colloid generation and, due to its high porosity, for colloid mobility. Colloid concentration measurements were carried out using an in-situ liquid particle counting system. The in-house developed counting system with three commercially available sensors allowed the detection of single particles and colloids at low concentrations in the size range 50-5000 nm. The counting system was tested using suspensions prepared from certified size standards. The concentrations of colloids with size range 50-1000 nm were measured in cement pore water, which was collected from a column filled with a highly permeable backfill mortar. The chemical composition of the pore water corresponded to a Ca(OH)2-controlled cement system. Colloid concentrations in the backfill pore water were found to be typically lower than approximately 0.1 ppm. The specific (geometric) surface areas of the colloid populations were in the range 240 m2 g(-1) to 770 m2 g(-1). The low colloid inventories observed in this study can be explained by the high ionic strength and Ca concentrations of the cement pore water. These conditions are favourable for colloid-colloid and colloid-backfill interactions and unfavourable for colloid-enhanced nuclide transport.

  3. Valorization of an industrial waste (sludge as an artificial pozzolan in cementitious materials

    Directory of Open Access Journals (Sweden)

    Sanae Lamrani

    2016-12-01

    Full Text Available The present study fits within the framework of sustainable management of sludge generated from wastewater treatment in industrial network. The studied sludge comes from an industry manufacturing sanitary ceramic products. Physical, chemical and mineralogical characterization was carried out in order to give an identity card to the sludge. We noted the absence of metal pollution. In order to evaluate its pozzolanic character, the industrial sludge has been subjected to thermal activation at various temperatures (from 650°C to 1000°C. The pozzolanic activity was evaluated by physico-chemical and mechanical methods. Pozzolanicity measurement by conductivity, Frattini and Chapelle Test revealed the existence of pozzolanic properties of calcined samples. The best pozzolanic reactivity was obtained for the sample calcined at 800°C. We noticed a decrease in the reactivity of the samples calcined from 850°C. In addition, analysis by means of X-ray diffraction and Fourier transform infrared spectroscopy showed that sludge recrystallization begins at a temperature of 850°C. Pozzolanicity index of the thermally treated samples was determined by measuring the mechanical resistance of mortar specimens previously kept in a saturated lime solution for 28 days. The best pozzolanic activity index was obtained for the sample calcined at 800°C (109.1%. The study of mechanical performances and resistance to chemical attacks of mortars incorporating sludge (calcined at 800°C with different percentages and at various ages showed an improvement of mechanical and chemical resistance compared to the control mortar (100% cement . This work is a contribution to the research for new supplying sources of raw materials and additives in the field of construction. It presents a proposition of a promising solution for the valorization of waste material as an additive instead of being discharged into open air dumps causing a major environmental problem.

  4. CERAMIC WASTE FORM DATA PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J.; Marra, J.

    2014-06-13

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

  5. Research and Development of a New Silica-Alumina Based Cementitious Material Largely Using Coal Refuse for Mine Backfill, Mine Sealing and Waste Disposal Stabilization

    Energy Technology Data Exchange (ETDEWEB)

    Henghu Sun; Yuan Yao

    2012-06-29

    Coal refuse and coal combustion byproducts as industrial solid waste stockpiles have become great threats to the environment. To activate coal refuse is one practical solution to recycle this huge amount of solid waste as substitute for Ordinary Portland Cement (OPC). The central goal of this project is to investigate and develop a new silica-alumina based cementitious material largely using coal refuse as a constituent that will be ideal for durable construction, mine backfill, mine sealing and waste disposal stabilization applications. This new material is an environment-friendly alternative to Ordinary Portland Cement. The main constituents of the new material are coal refuse and other coal wastes including coal sludge and coal combustion products (CCPs). Compared with conventional cement production, successful development of this new technology could potentially save energy and reduce greenhouse gas emissions, recycle vast amount of coal wastes, and significantly reduce production cost. A systematic research has been conducted to seek for an optimal solution for enhancing pozzolanic reactivity of the relatively inert solid waste-coal refuse in order to improve the utilization efficiency and economic benefit as a construction and building material.

  6. Degradation of cementitious materials associated with salstone disposal units

    Energy Technology Data Exchange (ETDEWEB)

    Flach, G. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith, F. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-01

    The Saltstone facilities at the DOE Savannah River Site (SRS) stabilize and dispose of low-level radioactive salt solution originating from liquid waste storage tanks at the site. The Saltstone Production Facility (SPF) receives treated salt solution and mixes the aqueous waste with dry cement, blast furnace slag, and fly ash to form a grout slurry which is mechanically pumped into concrete disposal cells that compose the Saltstone Disposal Facility (SDF). The solidified grout is termed “saltstone”. Cementitious materials play a prominent role in the design and long-term performance of the SDF. The saltstone grout exhibits low permeability and diffusivity, and thus represents a physical barrier to waste release. The waste form is also reducing, which creates a chemical barrier to waste release for certain key radionuclides, notably Tc-99. Similarly, the concrete shell of a saltstone disposal unit (SDU) represents an additional physical and chemical barrier to radionuclide release to the environment. Together the waste form and the SDU compose a robust containment structure at the time of facility closure. However, the physical and chemical state of cementitious materials will evolve over time through a variety of phenomena, leading to degraded barrier performance over Performance Assessment (PA) timescales of thousands to tens of thousands of years. Previous studies of cementitious material degradation in the context of low-level waste disposal have identified sulfate attack, carbonation influenced steel corrosion, and decalcification (primary constituent leaching) as the primary chemical degradation phenomena of most relevance to SRS exposure conditions. In this study, degradation time scales for each of these three degradation phenomena are estimated for saltstone and concrete associated with each SDU type under conservative, nominal, and best estimate assumptions.

  7. 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.

  8. Evaluation of blends bauxite-calcination-method red mud with other industrial wastes as a cementitious material: Properties and hydration characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Na, E-mail: zhangna06@mails.tsinghua.edu.cn [State Key Lab of New Ceramics and Fine Processing, Department of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Liu Xiaoming [School of Metallurgical and Ecological Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Sun Henghu [School of Engineering and Computer Science, University of the Pacific, Stockton, CA 95211 (United States); Li Longtu [State Key Lab of New Ceramics and Fine Processing, Department of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China)

    2011-01-15

    Red mud is generated from alumina production, and its disposal is currently a worldwide problem. In China, large quantities of red mud derived from bauxite calcination method are being discharged annually, and its utilization has been an urgent topic. This experimental research was to evaluate the feasibility of blends red mud derived from bauxite calcination method with other industrial wastes for use as a cementitious material. The developed cementitious material containing 30% of the bauxite-calcination-method red mud possessed compressive strength properties at a level similar to normal Portland cement, in the range of 45.3-49.5 MPa. Best compressive strength values were demonstrated by the specimen RSFC2 containing 30% bauxite-calcination-method red mud, 21% blast-furnace slag, 10% fly ash, 30% clinker, 8% gypsum and 1% compound agent. The mechanical and physical properties confirm the usefulness of RSFC2. The hydration characteristics of RSFC2 were characterized by XRD, FTIR, {sup 27}Al MAS-NMR and SEM. As predominant hydration products, ettringite and amorphous C-S-H gel are principally responsible for the strength development of RSFC2. Comparing with the traditional production for ordinary Portland cement, this green technology is easier to be implemented and energy saving. This paper provides a key solution to effectively utilize bauxite-calcination-method red mud.

  9. Evaluation of blends bauxite-calcination-method red mud with other industrial wastes as a cementitious material: properties and hydration characteristics.

    Science.gov (United States)

    Zhang, Na; Liu, Xiaoming; Sun, Henghu; Li, Longtu

    2011-01-15

    Red mud is generated from alumina production, and its disposal is currently a worldwide problem. In China, large quantities of red mud derived from bauxite calcination method are being discharged annually, and its utilization has been an urgent topic. This experimental research was to evaluate the feasibility of blends red mud derived from bauxite calcination method with other industrial wastes for use as a cementitious material. The developed cementitious material containing 30% of the bauxite-calcination-method red mud possessed compressive strength properties at a level similar to normal Portland cement, in the range of 45.3-49.5 MPa. Best compressive strength values were demonstrated by the specimen RSFC2 containing 30% bauxite-calcination-method red mud, 21% blast-furnace slag, 10% fly ash, 30% clinker, 8% gypsum and 1% compound agent. The mechanical and physical properties confirm the usefulness of RSFC2. The hydration characteristics of RSFC2 were characterized by XRD, FTIR, (27)Al MAS-NMR and SEM. As predominant hydration products, ettringite and amorphous C-S-H gel are principally responsible for the strength development of RSFC2. Comparing with the traditional production for ordinary Portland cement, this green technology is easier to be implemented and energy saving. This paper provides a key solution to effectively utilize bauxite-calcination-method red mud.

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

    Energy Technology Data Exchange (ETDEWEB)

    Randklev, E.H.

    1993-06-01

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

  11. 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

  12. Alternative solidified forms for nuclear wastes

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-01-01

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

  13. 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.

  14. Cementitious Barriers Partnership - FY2015 End-Year Report

    Energy Technology Data Exchange (ETDEWEB)

    Burns, H. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Flach, G. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Langton, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith, F. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kosson, D. S. [Vanderbilt Univ., Nashville, TN (United States). School of Engineering; Brown, K. G. [Vanderbilt Univ., Nashville, TN (United States). School of Engineering; Samson, E. [SIMCO Technologies, Inc., QC (Canada); Meeussen, J. C. L. [Nuclear Research and Consultancy Group (NRG); Seignette, Paul [Energy Research Center of the Netherlands; van der Sloot, H. A. [Hans van der Sloot Consultancy

    2015-09-17

    The DOE-EM Office of Tank Waste Management Cementitious Barriers Partnership (CBP) is chartered with providing the technical basis for implementing cement-based waste forms and radioactive waste containment structures for long-term disposal. Therefore, the CBP ultimate purpose is to support progress in final treatment and disposal of legacy waste and closure of High-Level Waste (HLW) tanks in the DOE complex. This status report highlights the CBP 2015 Software and Experimental Program efforts and accomplishments that support DOE needs in environmental cleanup and waste disposal. DOE needs in this area include: Long-term performance predictions to provide credibility (i.e., a defensible technical basis) for regulator and DOE review and approvals, Facility flow sheet development/enhancements, and Conceptual designs for new disposal facilities. In 2015, the CBP developed a beta release of the CBP Software Toolbox – “Version 3.0”, which includes new STADIUM carbonation and damage models, a new SRNL module for estimating hydraulic properties and flow in fractured and intact cementitious materials, and a new LeachXS/ORCHESTRA (LXO) oxidation module. In addition, the STADIUM sulfate attack and chloride models have been improved as well as the LXO modules for sulfate attack, carbonation, constituent leaching, and percolation with radial diffusion (for leaching and transport in cracked cementitious materials). These STADIUM and LXO models are applicable to and can be used by both DOE and the Nuclear Regulatory Commission (NRC) end-users for service life prediction and long-term leaching evaluations of radioactive waste containment structures across the DOE complex.

  15. Korean Waste Management Law and Waste Disposal Forms.

    Science.gov (United States)

    1991-03-01

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

  16. Combined Waste Form Cost Trade Study

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-11-01

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

  17. 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.

  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. 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...

  20. 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.

  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. Performance Test on Polymer Waste Form

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  3. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

  6. 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

  7. OVERVIEW OF THE U.S. DEPARTMENT OF ENERGY AND NUCLEAR REGULATORY COMMISSION PERFORMANCE ASSESSMENT APPROACHES: CEMENTITIOUS BARRIERS PARTNERSHIP

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Burns, H.

    2009-05-29

    Engineered barriers including cementitious barriers are used at sites disposing or contaminated with low-level radioactive waste to enhance performance of the natural environment with respect to controlling the potential spread of contaminants. Drivers for using cementitious barriers include: high radionuclide inventory, radionuclide characteristics (e.g., long half-live, high mobility due to chemical form/speciation, waste matrix properties, shallow water table, and humid climate that provides water for leaching the waste). This document comprises the first in a series of reports being prepared for the Cementitious Barriers Partnership. The document is divided into two parts which provide a summary of: (1) existing experience in the assessment of performance of cementitious materials used for radioactive waste management and disposal and (2) sensitivity and uncertainty analysis approaches that have been applied for assessments. Each chapter is organized into five parts: Introduction, Regulatory Considerations, Specific Examples, Summary of Modeling Approaches and Conclusions and Needs. The objective of the report is to provide perspective on the state of the practice for conducting assessments for facilities involving cementitious barriers and to identify opportunities for improvements to the existing approaches. Examples are provided in two contexts: (1) performance assessments conducted for waste disposal facilities and (2) performance assessment-like analyses (e.g., risk assessments) conducted under other regulatory regimes. The introductory sections of each section provide a perspective on the purpose of performance assessments and different roles of cementitious materials for radioactive waste management. Significant experience with assessments of cementitious materials associated with radioactive waste disposal concepts exists in the US Department of Energy Complex and the commercial nuclear sector. Recently, the desire to close legacy facilities has created

  8. 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

  9. 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.

  10. 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.

  11. Development of Alternative Technetium Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Czerwinski, Kenneth

    2013-09-13

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

  12. Properties of Calcium Acetate Manufactured with Etching Waste Solution and Limestone Sludge as a Cementitious High-Early-Strength Admixture

    Directory of Open Access Journals (Sweden)

    Deuck-Mo Kim

    2016-01-01

    Full Text Available Concrete is one of the most widely used construction materials. There are several methods available to improve its performance, with one of them being the use of high-early-strength admixtures (HESAs. Typical HESAs include calcium nitrate, calcium chloride, and calcium formate (CF. Industrial by-products, such as acetic acid and lime stone sludge (LSS, can be used together to produce calcium acetate (CA, which can subsequently be used as a cementitious HESA. In this study, calcium carbonate and LSS were mixed with cement in weight ratios of 1 : 1, 1 : 1.5, and 1 : 2, and the properties of the as-produced CA were evaluated. CA and CF were mixed with cement in different weight ratios (0, 1, 2, and 3 wt% to obtain CA- and CF-mortars, respectively. The flow behavior, setting time, pH, and compressive strength of these mortars were evaluated, and their X-ray diffraction patterns were also analyzed. It was found that as the CF content in the CF-mortar increased, the initial strength of the mortar also increased. However, it impaired its long-term strength. On the other hand, when 1% CA was mixed with cement, satisfactory early and long-term strengths were achieved. Thus, CA, which is obtained from industrial by-products, can be an effective HESA.

  13. 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

  14. Alternative Waste Forms for Electro-Chemical Salt Waste

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-28

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

  15. 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

    decrease the need for expensive engineered barriers.Our current work aims are 1) quantifying and understanding the processes associated with glass alteration in contact with Fe-bearing materials; 2) quantifying and understanding the processes associated with glass alteration in presence of MgO (example of engineered barrier used in WIPP); 3) identifying glass alteration suppressants and the processes involved to reach glass alteration suppression; 4) quantifying and understanding the processes associated with Saltstone and Cast Stone (SRS and Hanford cementitious waste forms) in various representative groundwaters; 5) investigating positron annihilation as a new tool for the study of glass alteration; and 6) quantifying and understanding the processes associated with glass alteration under gamma irradiation.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

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

    Science.gov (United States)

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

    2014-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-15

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

  2. Degradation Of Cementitious Materials Associated With Saltstone Disposal Units

    Energy Technology Data Exchange (ETDEWEB)

    Flach, G. P; Smith, F. G. III

    2013-03-19

    The Saltstone facilities at the DOE Savannah River Site (SRS) stabilize and dispose of low-level radioactive salt solution originating from liquid waste storage tanks at the site. The Saltstone Production Facility (SPF) receives treated salt solution and mixes the aqueous waste with dry cement, blast furnace slag, and fly ash to form a grout slurry which is mechanically pumped into concrete disposal cells that compose the Saltstone Disposal Facility (SDF). The solidified grout is termed “saltstone”. Cementitious materials play a prominent role in the design and long-term performance of the SDF. The saltstone grout exhibits low permeability and diffusivity, and thus represents a physical barrier to waste release. The waste form is also reducing, which creates a chemical barrier to waste release for certain key radionuclides, notably Tc-99. Similarly, the concrete shell of an SDF disposal unit (SDU) represents an additional physical and chemical barrier to radionuclide release to the environment. Together the waste form and the SDU compose a robust containment structure at the time of facility closure. However, the physical and chemical state of cementitious materials will evolve over time through a variety of phenomena, leading to degraded barrier performance over Performance Assessment (PA) timescales of thousands to tens of thousands of years. Previous studies of cementitious material degradation in the context of low-level waste disposal have identified sulfate attack, carbonation influenced steel corrosion, and decalcification (primary constituent leaching) as the primary chemical degradation phenomena of most relevance to SRS exposure conditions. In this study, degradation time scales for each of these three degradation phenomena are estimated for saltstone and concrete associated with each SDU type under conservative, nominal, and best estimate assumptions. The nominal value (NV) is an intermediate result that is more probable than the conservative

  3. Impact of cementitious materials decalcification on transfer properties: application to radioactive waste deep repository; Influence de la decalcification de materiaux cimentaires sur les proprietes de transfert: application au stockage profond de dechets radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    Perlot, C

    2005-09-15

    Cementitious materials have been selected to compose the engineering barrier system (EBS) of the French radioactive waste deep repository, because of concrete physico-chemical properties: the hydrates of the cementitious matrix and the pH of the pore solution contribute to radionuclides retention; furthermore the compactness of these materials limits elements transport. The confinement capacity of the system has to be assessed while a period at least equivalent to waste activity (up to 100.000 years). His durability was sustained by the evolution of transfer properties in accordance with cementitious materials decalcification, alteration that expresses structure long-term behavior. Then, two degradation modes were carried out, taking into account the different physical and chemical solicitations imposed by the host formation. The first mode, a static one, was an accelerated decalcification test using nitrate ammonium solution. It replicates the EBS alteration dues to underground water. Degradation kinetic was estimated by the amount of calcium leached and the measurement of the calcium hydroxide dissolution front. To evaluate the decalcification impact, samples were characterized before and after degradation in term of microstructure (porosity, pores size distribution) and of transfer properties (diffusivity, gas and water permeability). The influence of cement nature (ordinary Portland cement, blended cement) and aggregates type (lime or siliceous) was observed: experiments were repeated on different mortars mixes. On this occasion, an essential reflection on this test metrology was led. The second mode, a dynamical degradation, was performed with an environmental permeameter. It recreates the EBS solicitations ensured during the re-saturation period, distinguished by the hydraulic pressure imposed by the geologic layer and the waste exothermicity. This apparatus, based on triaxial cell functioning, allows applying on samples pressure drop between 2 and 10 MPa and

  4. 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

  5. Magnesium alloys and graphite wastes encapsulated in cementitious materials: Reduction of galvanic corrosion using alkali hydroxide activated blast furnace slag.

    Science.gov (United States)

    Chartier, D; Muzeau, B; Stefan, L; Sanchez-Canet, J; Monguillon, C

    2017-03-15

    Magnesium alloys and graphite from spent nuclear fuel have been stored together in La Hague plant. The packaging of these wastes is under consideration. These wastes could be mixed in a grout composed of industrially available cement (Portland, calcium aluminate…). Within the alkaline pore solution of these matrixes, magnesium alloys are imperfectly protected by a layer of Brucite resulting in a slow corrosion releasing hydrogen. As the production of this gas must be considered for the storage safety, and the quality of wasteform, it is important to select a cement matrix capable of lowering the corrosion kinetics. Many types of calcium based cements have been tested and most of them have caused strong hydrogen production when magnesium alloys and graphite are conditioned together because of galvanic corrosion. Exceptions are binders based on alkali hydroxide activated ground granulated blast furnace slag (BFS) which are presented in this article. Copyright © 2016 Elsevier B.V. All rights reserved.

  6. Intended long term performances of cementitious engineered barriers for future storage and disposal facilities for radioactive wastes in Romania

    OpenAIRE

    Sociu F.; Georgescu R.; Toma R.; Barariu Gh.; Fako R.

    2013-01-01

    Considering the EU statements, Romania is engaged to endorse in the near future the IAEA relevant publications on geological repository (CNCANa), to update the Medium and Long Term National Strategy for Safe Management of Radioactive Waste and to approve the Road Map for Geological Repository Development. Currently, for example, spent fuel is wet stored for 6 years and after this period it is transported to dry storage in MACSTOR-200 (a concrete monolithic module) where it is intended to rema...

  7. Magnesium alloys and graphite wastes encapsulated in cementitious materials: Reduction of galvanic corrosion using alkali hydroxide activated blast furnace slag

    Energy Technology Data Exchange (ETDEWEB)

    Chartier, D., E-mail: david.chartier@cea.fr [Commissariat à l' Energie Atomique et aux Energies Alternatives, CEA, DEN, DTCD, SPDE, F-30207 Bagnols-sur-Cèze (France); Muzeau, B. [DEN-Service d’Etude du Comportement des Radionucléides (SECR), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Stefan, L. [AREVA NC/D& S - France/Technical Department, 1 place Jean Millier 92084 Paris La Défense (France); Sanchez-Canet, J. [Commissariat à l' Energie Atomique et aux Energies Alternatives, CEA, DEN, DTCD, SPDE, F-30207 Bagnols-sur-Cèze (France); Monguillon, C. [DEN-Service d’Etude du Comportement des Radionucléides (SECR), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2017-03-15

    Highlights: • Embedded in cement, magnesium is corroded by residual water present in porosity of the matrix. • Corrosion is enhanced by galvanic phenomenon when magnesium is in contact with graphite. • Galvanic corrosion of magnesium in contact with graphite debris is shown to be severe with ordinary Portland cement. • Galvanic corrosion is significantly lowered in high alkali medium such as sodium hydroxide. • Sodium hydroxide activated blast furnace slag is a convenient binder to embed magnesium. - Abstract: Magnesium alloys and graphite from spent nuclear fuel have been stored together in La Hague plant. The packaging of these wastes is under consideration. These wastes could be mixed in a grout composed of industrially available cement (Portland, calcium aluminate…). Within the alkaline pore solution of these matrixes, magnesium alloys are imperfectly protected by a layer of Brucite resulting in a slow corrosion releasing hydrogen. As the production of this gas must be considered for the storage safety, and the quality of wasteform, it is important to select a cement matrix capable of lowering the corrosion kinetics. Many types of calcium based cements have been tested and most of them have caused strong hydrogen production when magnesium alloys and graphite are conditioned together because of galvanic corrosion. Exceptions are binders based on alkali hydroxide activated ground granulated blast furnace slag (BFS) which are presented in this article.

  8. 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.

  9. Accelerated leach test of paraffin waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  10. Electrochemical corrosion testing of metal waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-14

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

  11. Monazite as a suitable actinide waste form

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-03-01

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

  12. 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.

  13. Safeguards and retrievability from waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Danker, W.

    1996-05-01

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

  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. 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.

  16. 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.

  17. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-31

    The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). The pretreatment facility will have the capacity to separate all of the tank wastes into the HLW and LAW fractions, and the HLW Vitrification Facility will have the capacity to vitrify all of the HLW. However, a second immobilization facility will be needed for the expected volume of LAW requiring immobilization. A number of alternatives, including Cast Stone—a cementitious waste form—are being considered to provide the additional LAW immobilization capacity.

  20. VARIABILITY OF KD VALUES IN CEMENTITIOUS MATERIALS AND SEDIMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Almond, P.; Kaplan, D.; Shine, E.

    2012-02-02

    Measured distribution coefficients (K{sub d} values) for environmental contaminants provide input data for performance assessments (PA) that evaluate physical and chemical phenomena for release of radionuclides from wasteforms, degradation of engineered components and subsequent transport of radionuclides through environmental media. Research efforts at SRNL to study the effects of formulation and curing variability on the physiochemical properties of the saltstone wasteform produced at the Saltstone Disposal Facility (SDF) are ongoing and provide information for the PA and Saltstone Operations. Furthermore, the range and distribution of plutonium K{sub d} values in soils is not known. Knowledge of these parameters is needed to provide guidance for stochastic modeling in the PA. Under the current SRS liquid waste processing system, supernate from F & H Tank Farm tanks is processed to remove actinides and fission products, resulting in a low-curie Decontaminated Salt Solution (DSS). At the Saltstone Production Facility (SPF), DSS is mixed with premix, comprised of blast furnace slag (BFS), Class F fly ash (FA), and portland cement (OPC) to form a grout mixture. The fresh grout is subsequently placed in SDF vaults where it cures through hydration reactions to produce saltstone, a hardened monolithic waste form. Variation in saltstone composition and cure conditions of grout can affect the saltstone's physiochemical properties. Variations in properties may originate from variables in DSS, premix, and water to premix ratio, grout mixing, placing, and curing conditions including time and temperature (Harbour et al. 2007; Harbour et al. 2009). There are no previous studies reported in the literature regarding the range and distribution of K{sub d} values in cementitious materials. Presently, the Savannah River Site (SRS) estimate ranges and distributions of K{sub d} values based on measurements of K{sub d} values made in sandy SRS sediments (Kaplan 2010). The actual

  1. 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

  2. Final report on cermet high-level waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-08-01

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

  3. 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

  4. 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.

  5. 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

  6. Minerals as natural analogues for crystalline nuclear waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

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

    Science.gov (United States)

    Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

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

  9. 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.

  10. 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.

  11. Weathering Effects on Technetium Leachability from Ceramicrete Waste Form

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-03-01

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

  13. DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

    Energy Technology Data Exchange (ETDEWEB)

    J. CUNNANE

    2004-11-19

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

  14. 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

  15. Timing of Getter Material Addition in Cementitious Wasteforms

    Science.gov (United States)

    Lawter, A.; Qafoku, N. P.; Asmussen, M.; Neeway, J.; Smith, G. L.

    2015-12-01

    A cementitious waste form, Cast Stone, is being evaluated as a possible supplemental immobilization technology for the Hanford sites's low activity waste (LAW), which contains radioactive 99Tc and 129I, as part of the tank waste cleanup mission. Cast Stone is made of a dry blend 47% blast furnace slag, 45% fly ash, and 8% ordinary Portland cement, mixed with a low-activity waste (LAW). To improve the retention of Tc and/or I in Cast Stone, materials with a high affinity for Tc and/or I, termed "getters," can be added to provide a stable domain for the radionuclides of concern. Previous testing conducted with a variety of getters has identified Tin(II)-Apatite and Silver Exchanged Zeolite as promising candidates for Tc and I, respectively. Investigation into the sequence in which getters are added to Cast Stone was performed following two methods: 1) adding getters to the Cast Stone dry blend, and then mixing with liquid waste, and 2) adding getters to the liquid waste first, followed by addition of the Cast Stone dry blend. Cast Stone monolith samples were prepared with each method and leach tests, following EPA method 1315, were conducted in either distilled water or simulated vadose zone porewater for a period of up to 63 days. The leachate was analyzed for Tc, I, Na, NO3-, NO2- and Cr with ICP-MS, ICP-OES and ion chromatography and the results indicated that the Cast Stone with getter addition in the dry blend mix (method 1) has lower rates of Tc and I leaching. The mechanisms of radionuclide release from the Cast Stone were also investigated with a variety of solid phase characterization techniques of the monoliths before and after leaching, such as XRD, SEM/EDS, TEM/SAED and other spectroscopic techniques.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-09-01

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

  17. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-23

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

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

    Science.gov (United States)

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

    2014-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  1. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-02-10

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

  3. 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.

  4. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Alexandra Navrotsky

    2010-07-30

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

  6. Experimental Study on Cementitious Composites Embedded with Organic Microcapsules

    NARCIS (Netherlands)

    Wang, X.; Xing, F.; Zhang, M.; Han, N.; Qian, Z.

    2013-01-01

    The recovery behavior for strength and impermeability of cementitious composites embedded with organic microcapsules was investigated in this study. Mortar specimens were formed by mixing the organic microcapsules and a catalyst with cement and sand. The mechanical behaviors of flexural and compress

  7. Technetium Waste Form Development - Progress Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-07

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-01

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

  11. 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.

  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. Advanced waste forms research and development. Annual report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, G.J.

    1975-06-11

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

  14. Self-degradable Cementitious Sealing Materials

    Energy Technology Data Exchange (ETDEWEB)

    Sugama, T.; Butcher, T., Lance Brothers, Bour, D.

    2010-10-01

    A self-degradable alkali-activated cementitious material consisting of a sodium silicate activator, slag, Class C fly ash, and sodium carboxymethyl cellulose (CMC) additive was formulated as one dry mix component, and we evaluated its potential in laboratory for use as a temporary sealing material for Enhanced Geothermal System (EGS) wells. The self-degradation of alkali-activated cementitious material (AACM) occurred, when AACM heated at temperatures of {ge}200 C came in contact with water. We interpreted the mechanism of this water-initiated self-degradation as resulting from the in-situ exothermic reactions between the reactants yielded from the dissolution of the non-reacted or partially reacted sodium silicate activator and the thermal degradation of the CMC. The magnitude of self-degradation depended on the CMC content; its effective content in promoting degradation was {ge}0.7%. In contrast, no self-degradation was observed from CMC-modified Class G well cement. For 200 C-autoclaved AACMs without CMC, followed by heating at temperatures up to 300 C, they had a compressive strength ranging from 5982 to 4945 psi, which is {approx}3.5-fold higher than that of the commercial Class G well cement; the initial- and final-setting times of this AACM slurry at 85 C were {approx}60 and {approx}90 min. Two well-formed crystalline hydration phases, 1.1 nm tobermorite and calcium silicate hydrate (I), were responsible for developing this excellent high compressive strength. Although CMC is an attractive, as a degradation-promoting additive, its addition to both the AACM and the Class G well cement altered some properties of original cementitious materials; among those were an extending their setting times, an increasing their porosity, and lowering their compressive strength. Nevertheless, a 0.7% CMC-modified AACM as self-degradable cementitious material displayed the following properties before its breakdown by water; {approx}120 min initial- and {approx}180 min final

  15. Degradation modeling of the ANL ceramic waste form

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-28

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

  16. Electrochemical migration technique to accelerate ageing of cementitious materials

    Directory of Open Access Journals (Sweden)

    Abbas Z.

    2013-07-01

    Full Text Available Durability assessment of concrete structures for constructions in nuclear waste repositories requires long term service life predictions. As deposition of low and intermediate level radioactive waste (LILW takes up to 100 000 years, it is necessary to analyze the service life of cementitious materials in this time perspective. Using acceleration methods producing aged specimens would decrease the need of extrapolating short term data sets. Laboratory methods are therefore, needed for accelerating the ageing process without making any influencing distortion in the properties of the materials. This paper presents an electro-chemical migration method to increase the rate of calcium leaching from cementitious specimens. This method is developed based on the fact that major long term deterioration process of hardened cement paste in concrete structures for deposition of LILW is due to slow diffusion of calcium ions. In this method the cementitious specimen is placed in an electrochemical cell as a porous path way through which ions can migrate at a rate far higher than diffusion process. The electrical field is applied to the cell in a way to accelerate the ion migration without making destructions in the specimen’s micro and macroscopic properties. The anolyte and catholyte solutions are designed favoring dissolution of calcium hydroxide and compensating for the leached calcium ions with another ion like lithium.

  17. Electrochemical migration technique to accelerate ageing of cementitious materials

    Science.gov (United States)

    Babaahmadi, A.; Tang, L.; Abbas, Z.

    2013-07-01

    Durability assessment of concrete structures for constructions in nuclear waste repositories requires long term service life predictions. As deposition of low and intermediate level radioactive waste (LILW) takes up to 100 000 years, it is necessary to analyze the service life of cementitious materials in this time perspective. Using acceleration methods producing aged specimens would decrease the need of extrapolating short term data sets. Laboratory methods are therefore, needed for accelerating the ageing process without making any influencing distortion in the properties of the materials. This paper presents an electro-chemical migration method to increase the rate of calcium leaching from cementitious specimens. This method is developed based on the fact that major long term deterioration process of hardened cement paste in concrete structures for deposition of LILW is due to slow diffusion of calcium ions. In this method the cementitious specimen is placed in an electrochemical cell as a porous path way through which ions can migrate at a rate far higher than diffusion process. The electrical field is applied to the cell in a way to accelerate the ion migration without making destructions in the specimen's micro and macroscopic properties. The anolyte and catholyte solutions are designed favoring dissolution of calcium hydroxide and compensating for the leached calcium ions with another ion like lithium.

  18. Concrete and cement composites used for radioactive waste deposition.

    Science.gov (United States)

    Koťátková, Jaroslava; Zatloukal, Jan; Reiterman, Pavel; Kolář, Karel

    2017-08-23

    This review article presents the current state-of-knowledge of the use of cementitious materials for radioactive waste disposal. An overview of radwaste management processes with respect to the classification of the waste type is given. The application of cementitious materials for waste disposal is divided into two main lines: i) as a matrix for direct immobilization of treated waste form; and ii) as an engineered barrier of secondary protection in the form of concrete or grout. In the first part the immobilization mechanisms of the waste by cement hydration products is briefly described and an up-to date knowledge about the performance of different cementitious materials is given, including both traditional cements and alternative binder systems. The advantages, disadvantages as well as gaps in the base of information in relation to individual materials are stated. The following part of the article is aimed at description of multi-barrier systems for intermediate level waste repositories. It provides examples of proposed concepts by countries with advanced waste management programmes. In the paper summary, the good knowledge of the material durability due to its vast experience from civil engineering is highlighted however with the urge for specific approach during design and construction of a repository in terms of stringent safety requirements. Copyright © 2017. Published by Elsevier Ltd.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-15

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

  20. Cementitious Barriers Partnership FY2013 End-Year Report

    Energy Technology Data Exchange (ETDEWEB)

    Flach, G. P. [Savannah River Site (SRS), Aiken, SC (United States); Langton, C. A. [Savannah River Site (SRS), Aiken, SC (United States); Burns, H. H. [Savannah River Site (SRS), Aiken, SC (United States); Smith, F. G. [Savannah River Site (SRS), Aiken, SC (United States); Kosson, D. S. [Vanderbilt University, School of Engineering, Nashville, TN (United States); Brown, K. G. [Vanderbilt University, School of Engineering, Nashville, TN (United States); Samson, E. [SIMCO Technologies, Inc., Quebec (Canada); Meeussen, J. C.L. [Nuclear Research and Consultancy Group (NRG), Petten (The Netherlands); van der Sloot, H. A. [Hans van der Sloot Consultancy, Langedijk (The Netherlands); Garboczi, E. J. [Materials & Construction Research Division, National Institute of Standards and Technology, Gaithersburg, MD (United States)

    2013-11-01

    hydraulic and constituent mass transfer parameters needed in modeling. Two CBP software demonstrations were conducted in FY2013, one to support the Saltstone Disposal Facility (SDF) at SRS and the other on a representative Hanford high-level waste tank. The CBP Toolbox demonstration on the SDF provided analysis on the most probable degradation mechanisms to the cementitious vault enclosure caused by sulfate and carbonation ingress. This analysis was documented and resulted in the issuance of a SDF Performance Assessment Special Analysis by Liquid Waste Operations this fiscal year. The two new software tools supporting chloride attack and dual-regime flow will provide additional degradation tools to better evaluate performance of DOE and commercial cementitious barriers. The CBP SRNL experimental program produced two patent applications and field data that will be used in the development and calibration of CBP software tools being developed in FY2014. The CBP software and simulation tools varies from other efforts in that all the tools are based upon specific and relevant experimental research of cementitious materials utilized in DOE applications. The CBP FY2013 program involved continuing research to improve and enhance the simulation tools as well as developing new tools that model other key degradation phenomena not addressed in Version 1.0. Also efforts to continue to verify the various simulation tools through laboratory experiments and analysis of field specimens are ongoing and will continue into FY2014 to quantify and reduce the uncertainty associated with performance assessments. This end-year report summarizes FY2013 software development efforts and the various experimental programs that are providing data for calibration and validation of the CBP developed software.

  1. 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.

  2. 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

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

    Science.gov (United States)

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

    2016-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-10-11

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

  5. 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.

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

    Science.gov (United States)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-02-25

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

  8. REVIEW OF MECHANISTIC UNDERSTANDING AND MODELING AND UNCERTAINTY ANALYSIS METHODS FOR PREDICTING CEMENTITIOUS BARRIER PERFORMANCE

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Kosson, D.

    2009-11-30

    Cementitious barriers for nuclear applications are one of the primary controls for preventing or limiting radionuclide release into the environment. At the present time, performance and risk assessments do not fully incorporate the effectiveness of engineered barriers because the processes that influence performance are coupled and complicated. Better understanding the behavior of cementitious barriers is necessary to evaluate and improve the design of materials and structures used for radioactive waste containment, life extension of current nuclear facilities, and design of future nuclear facilities, including those needed for nuclear fuel storage and processing, nuclear power production and waste management. The focus of the Cementitious Barriers Partnership (CBP) literature review is to document the current level of knowledge with respect to: (1) mechanisms and processes that directly influence the performance of cementitious materials (2) methodologies for modeling the performance of these mechanisms and processes and (3) approaches to addressing and quantifying uncertainties associated with performance predictions. This will serve as an important reference document for the professional community responsible for the design and performance assessment of cementitious materials in nuclear applications. This review also provides a multi-disciplinary foundation for identification, research, development and demonstration of improvements in conceptual understanding, measurements and performance modeling that would be lead to significant reductions in the uncertainties and improved confidence in the estimating the long-term performance of cementitious materials in nuclear applications. This report identifies: (1) technology gaps that may be filled by the CBP project and also (2) information and computational methods that are in currently being applied in related fields but have not yet been incorporated into performance assessments of cementitious barriers. The various

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

    Science.gov (United States)

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

    2017-02-01

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

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pegg, I.; Gould, T.

    1996-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ken Bateman; Stephen Priebe

    2006-08-01

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

  14. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-01

    Chemically bonded phosphate ceramics (CBPCs) have several advantages that make them ideal candidates for containing radioactive and hazardous wastes. In general, phosphates have high solid-solution capacities for incorporating radionuclides, as evidenced by several phosphates (e.g., monazites and apatites) that are natural analogs of radioactive and rare-earth elements. The phosphates have high radiation stability, are refractory, and will not degrade in the presence of internal heating by fission products. Dense and hard CBPCs can be fabricated inexpensively and at low temperature by acid-base reactions between an inorganic oxide/hydroxide powder and either phosphoric acid or an acid-phosphate solution. The resulting phosphates are extremely insoluble in aqueous media and have excellent long-term durability. CBPCs offer the dual stabilization mechanisms of chemical fixation and physical encapsulation, resulting in superior waste forms. The goal of this task is develop and demonstrate the feasibility of CBPCs for S/S of wastes containing fission products. The focus of this work is to develop a low-temperature CBPC immobilization system for eluted {sup 99}Tc wastes from sorption processes.

  15. Description of DWPF reference waste form and canister

    Energy Technology Data Exchange (ETDEWEB)

    1981-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Finkeldei, Sarah Charlotte

    2015-07-01

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

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

    Science.gov (United States)

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

    2016-12-01

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

  18. Radiation damage studies related to nuclear waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-12-01

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

  19. 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.

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

    Science.gov (United States)

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

    1992-01-01

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

  1. 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.

  2. Solidifications/stabilization treatability study of a mixed waste sludge

    Energy Technology Data Exchange (ETDEWEB)

    Spence, R.D. [Oak Ridge National Lab., TN (United States); Stine, E.F. [International Technologies Corp., Knoxville, TN (United States). Technology Development Lab.

    1996-03-01

    The Department of Energy Oak Ridge Operations Office signed a Federal Facility Compliance Agreement with the US Environmental Protection Agency Region IV regarding mixed wastes from the Oak Ridge Reservation (ORR) subject to the land disposal restriction provisions of the Resource Conservation and Recovery Act (RCRA). This agreement required treatability studies of solidification/stabilization (S/S) on mixed wastes from the ORR. This paper reports the results of the cementitious S/S studies conducted on a waste water treatment sludge generated from biodenitrification and heavy metals precipitation. For the cementitious waste forms, the additives tested were Portland cement, ground granulated blast furnace slag, Class F fly ash, and perlite. The properties measured on the treated waste were density, free-standing liquid, unconfined compressive strength, and TCLP performance. Spiking up to 10,000, 10,000, and 4,400 mg/kg of nickel, lead, and cadmium, respectively, was conducted to test waste composition variability and the stabilization limitations of the binding agents. The results indicated that nickel, lead and cadmium were stabilized fairly well in the high pH hydroxide-carbonate- ``bug bones`` sludge, but also clearly confirmed the established stabilization potential of cementitious S/S for these RCRA metals.

  3. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  5. 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.

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

    Science.gov (United States)

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

    1995-02-01

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

  7. Garnet nuclear waste forms – Solubility at repository conditions

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-15

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

  8. Radiation damage of hollandite in multiphase ceramic waste forms

    Science.gov (United States)

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

    2017-10-01

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

  9. 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.

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

    Science.gov (United States)

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

    2012-01-01

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

  11. Concrete mixture characterization. Cementitious barriers partnership

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Protiere, Yannick [SIMCO Technologies, Inc., Quebec (Canada)

    2014-12-01

    This report summarizes the characterization study performed on two concrete mixtures used for radioactive waste storage. Both mixtures were prepared with approximately 425 kg of binder. The testing protocol mostly focused on determining the transport properties of the mixtures; volume of permeable voids (porosity), diffusion coefficients, and water permeability were evaluated. Tests were performed after different curing durations. In order to obtain data on the statistical distribution of transport properties, the measurements after 2 years of curing were performed on 10+ samples. Overall, both mixtures exhibited very low tortuosities and permeabilities, a direct consequence of their low water-to-binder ratio and the use of supplementary cementitious materials. The data generated on 2-year old samples showed that porosity, tortuosity and permeability follow a normal distribution. Chloride ponding tests were also performed on test samples. They showed limited chloride ingress, in line with measured transport properties. These test results also showed that both materials react differently with chloride, a consequence of the differences in the binder chemical compositions.

  12. Cementitious barriers partnership concrete mixture characterization

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Protiere, Yannick [SIMCO Technologies, Inc., Quebec (Canada)

    2014-12-01

    This report summarizes the characterization study performed on two concrete mixtures used for radioactive waste storage. Both mixtures were prepared with approximately 425 kg of binder. The testing protocol mostly focused on determining the transport properties of the mixtures; volume of permeable voids (porosity), diffusion coefficients, and water permeability were evaluated. Tests were performed after different curing durations. In order to obtain data on the statistical distribution of transport properties, the measurements after 2 years of curing were performed on 10+ samples. Overall, both mixtures exhibited very low tortuosities and permeabilities, a direct consequence of their low water-to-binder ratio and the use of supplementary cementitious materials. The data generated on 2-year old samples showed that porosity, tortuosity and permeability follow a normal distribution. Chloride ponding tests were also performed on test samples. They showed limited chloride ingress, in line with measured transport properties. These test results also showed that both materials react differently with chloride, a consequence of the differences in the binder chemical compositions.

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

    Science.gov (United States)

    Harkins, Devin J. H.

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-30

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

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

    Science.gov (United States)

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

    2016-06-01

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

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Charles W. Solbrig; Kenneth J. Bateman

    2010-11-01

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

  18. 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.

  19. Progress in forming bottom barriers under waste sites

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  20. Waste form characteristics report, revision 1.3

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  1. Microscopic characterization of crystalline phases in waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  2. Service life prediction and cementitious composites

    DEFF Research Database (Denmark)

    Stoklund Larsen, E.

    The present Ph.D.thesis describes and discusses the applicability of a systematic methodology recommended by CIB W80/RILEM-PSL for sevice life prediction. The report describes the most important inherent and environmental factors affecting the service life of structures of cementitious composites....... On the basis of this discription of factors and experience from a test programme described in SBI Report 222, Service life prediction and fibre reinforced cementitious composites, the applicabillity of the CIB/RILEM methodology is discussed....

  3. 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.

  4. 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.

  5. Fracture propagation in cementitious materials

    DEFF Research Database (Denmark)

    Skocek, Jan

    is put on phenomena of a similar scale as is the characteristic size of inhomogeneities of the material, a model which re ects the heterogeneous nature of the material needs to be applied. This is, indeed, the case for prediction of mechanical properties of a material based on the knowledge of properties...... of the materials needs to be able to take this complexity into account. In this thesis, two frameworks for prediction of strengths of cementitious materials are developed. The rst one relates the strength of materials with aggregates with the properties of the matrix and distribution of aggregates. The second one...... relates the strength of cement paste with the properties of cement phases and its microstructure. The frameworks consist of an experimental part, an identication of material properties from the experiments and a modeling part based on an approximative discrete particle model. In the case of mortar...

  6. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  7. LONG-TERM TECHNETIUM INTERACTIONS WITH REDUCING CEMENTITIOUS MATERIALS

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, D.; Lilley, M.; Almond, P.; Powell, B.

    2011-03-15

    Technetium is among the key risk drivers at the Saltstone Facility. The way that it is immobilized in this cementitious waste form is by converting its highly mobile Tc(VII) form to a much less mobile Tc(IV) form through reduction by the cement's blast furnace slag. This report includes a review of published data and experimental results dealing with Tc leaching from Portland cement waste forms. The objectives for the literature study were to document previous reports of Tc interactions with slag-containing cementitious materials. The objectives for the laboratory study were to measure Tc-saltstone Kd values under reducing conditions. From the literature it was concluded: (1) Spectroscopic evidence showed that when Tc(IV) in a slag-cement was exposed to an oxidizing environment, it will convert to the more mobile Tc(VII) species within a short time frame, 2.5 years. (2) SRS saltstone will reduce Tc(VII) in the absence of NaS or sodium dithionite in a reducing atmosphere. (3) Only trace concentrations of atmospheric oxygen (30 to 60 ppm O{sub 2}; Eh 120 mV) at the high pH levels of cementitious systems is required to maintain Tc as Tc(VII). (4) Experimental conditions must be responsible for wide variability of measured K{sub d} values, such that they are either very low, {approx}1 mL/g, or they are very high {approx}1000 mL/g, suggesting that Tc(VII) or Tc(IV) dominate the systems. Much of this variability appears to be the result of experimental conditions, especially direct controls of oxygen contact with the sample. (5) A field study conducted at SRS in the 1980s indicated that a slag-saltstone immobilized Tc for 2.5 years. Below background concentrations of Tc leached out of the slag-containing saltstone, whereas Tc leached out of the slag-free saltstone at the rate of nitrate loss. One possible explanation for the immobilization of Tc in this study was that the slag-saltstone maintained reducing conditions within the core of the 55-gallon sample, whereas

  8. 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

  9. Naturally occurring glasses: analogues for radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-04-01

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

  10. Naturally occurring glasses: analogues for radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-04-01

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

  11. 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

  12. a Methodology for Thermal Characterization of Cementitious Materials.

    Science.gov (United States)

    Rousan, Akram Abdel-Majeed

    This study was an attempt to develop and utilize a method to measure thermal conductivity of cementitious materials without affecting the state of equilibrium of the sample. The thermal comparator method employed and developed here was proved to be suitable for this kind of material. The method was used to measure thermal conductivity of neat cement and cement plus additives. An attempt was also made to study other thermal properties of cementitious materials, such as thermal expansion and early heat of hydration. Class C portland cement was used in this study, mixed with different water to cement ratios and with additives and admixtures (fine quartz, fly ash, and quartz and basalt sands); and the thermal properties were measured at curing times up to three months. Thermal conductivity and thermal expansion were found to be dependent upon curing times, moisture contents, and concentrations and types of admixed phases. Mathematical relations of thermal conductivity as a function of additive contents were obtained at different curing times for the additives used in this study. Both thermal conductivity and thermal expansion were found to reach constant values when cementitious materials approach maturity. It was shown that thermal conductivity of mature hydrated additives can be estimated using the experimental results from measurements on composites. The hydration rate was also studied and the effect of additives as accelerators or retarders was explored using the rate of heat evolution curves. Results indicate that inert additives like quartz have no significant effect on the hydration rate, and reactive additives like fly ash tend, in general, to accelerate the reaction. Applications of the methods to the determination of properties of seal materials for geologic repositories for nuclear waste are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-12-01

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

  14. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-07-01

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

  16. 40 CFR 761.205 - Notification of PCB waste activity (EPA Form 7710-53).

    Science.gov (United States)

    2010-07-01

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

  17. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-04-01

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

  19. 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.

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

    OpenAIRE

    Carmalin Sophia Ayyappan

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

  2. 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).

  3. 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.

  4. Transport properties of damaged materials. Cementitious barriers partnership

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-01

    The objective of the Cementitious Barriers Partnership (CBP) project is to develop tools to improve understanding and prediction of the long-term structural, hydraulic, and chemical performance of cementitious barriers used in low-level waste storage applications. One key concern for the long-term durability of concrete is the degradation of the cementitious matrix, which occurs as a result of aggressive chemical species entering the material or leaching out in the environment, depending on the exposure conditions. The objective of the experimental study described in this report is to provide experimental data relating damage in cementitious materials to changes in transport properties, which can eventually be used to support predictive model development. In order to get results within a reasonable timeframe and to induce as much as possible uniform damage level in materials, concrete samples were exposed to freezing and thawing (F/T) cycles. The methodology consisted in exposing samples to F/T cycles and monitoring damage level with ultrasonic pulse velocity measurements. Upon reaching pre-selected damage levels, samples were tested to evaluate changes in transport properties. Material selection for the study was motivated by the need to get results rapidly, in order to assess the relevance of the methodology. Consequently, samples already available at SIMCO from past studies were used. They consisted in three different concrete mixtures cured for five years in wet conditions. The mixtures had water-to-cement ratios of 0.5, 0.65 and 0.75 and were prepared with ASTM Type I cement only. The results showed that porosity is not a good indicator for damage caused by the formation of microcracks. Some materials exhibited little variations in porosity even for high damage levels. On the other hand, significant variations in tortuosity were measured in all materials. This implies that damage caused by internal pressure does not necessarily create additional pore space in

  5. Cementitious barriers partnership transport properties of damaged materials

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-01

    The objective of the Cementitious Barriers Partnership (CBP) project is to develop tools to improve understanding and prediction of the long-term structural, hydraulic, and chemical performance of cementitious barriers used in low level waste storage applications. One key concern for the long-term durability of concrete is the degradation of the cementitious matrix, which occurs as a result of aggressive chemical species entering the material or leaching out in the environment, depending on the exposure conditions. The objective of the experimental study described in this report is to provide experimental data relating damage in cementitious materials to changes in transport properties, which can eventually be used to support predictive model development. In order to get results within a reasonable timeframe and to induce as much as possible uniform damage level in materials, concrete samples were exposed to freezing and thawing (F/T) cycles. The methodology consisted in exposing samples to F/T cycles and monitoring damage level with ultrasonic pulse velocity measurements. Upon reaching pre-selected damage levels, samples were tested to evaluate changes in transport properties. Material selection for the study was motivated by the need to get results rapidly, in order to assess the relevance of the methodology. Consequently, samples already available at SIMCO from past studies were used. They consisted in three different concrete mixtures cured for five years in wet conditions. The mixtures had water-to-cement ratios of 0.5, 0.65 and 0.75 and were prepared with ASTM Type I cement only. The results showed that porosity is not a good indicator for damage caused by the formation of microcracks. Some materials exhibited little variations in porosity even for high damage levels. On the other hand, significant variations in tortuosity were measured in all materials. This implies that damage caused by internal pressure do not necessarily creates additional pore space in

  6. Cracks and pores - Their roles in the transmission of water confined in cementitious materials

    Science.gov (United States)

    Bordallo, H. N.; Aldridge, L. P.; Wuttke, J.; Fernando, K.; Bertram, W. K.; Pardo, L. C.

    2010-10-01

    Cement paste is formed through a process called hydration by combining water with a cementitious material. Concrete, the worlds most versatile and most widely used material, can then be obtained when aggregates (sand, gravel, crushed stone) are added to the paste. The quality of hardened concrete is greatly influenced by the water confined in the cementitious materials and how it is transmitted through cracks and pores. Here we demonstrate that the water transport in cracks and capillary pores of hardened cement pastes can be approximately modeled by simple equations. Our findings highlight the significance of arresting the development of cracks in cementitious materials used in repository barriers. We also show that neutron scattering is an advantageous technique for understanding how water transmission is effected by gel pore structures. Defining measurable differences in gel pores may hold a key to prediction of the reduction of water transport through cement barriers.

  7. Treated Coconut Coir Pith as Component of Cementitious Materials

    Directory of Open Access Journals (Sweden)

    Dana Koňáková

    2015-01-01

    Full Text Available The presented paper deals with utilization of raw and treated coir pith as potential component of cementitious composites. The studied material is coir pith originating from a coconut production. Its applicability as cement mixture component was assessed in terms of the physical properties of concrete containing different amount of coir pith. Basic physical properties, compressive and bending strength, and hygric transport characteristics as well as thermal properties belong among the studied characteristics. It was proved that the concrete with 5% (by mass of cement of this waste material shows appropriate physical properties and it gives rise to an applicable material for building structures. Generally, the coir pith can be regarded as lightening additive. When 10% of coir pith was added, it has led to higher deterioration of properties than what is acceptable since such dosing is greatly increasing the total porosity. The influence of chemical treatment of coir pith was evaluated as well; both tested treatment methods improved the performance of cementitious composites while the acetylation was somewhat more effective the treatment by NaOH.

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

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, R.G.

    1983-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lindle, Dennis W.

    2011-04-21

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

  10. 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).

  11. 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).

  12. 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.

  13. 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.

  14. 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.

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

    Science.gov (United States)

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

    2016-05-01

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

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    H.W> Stockman; S. LeStrange

    2000-09-28

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-02-15

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

  19. 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.

  20. CEMENTITIOUS BARRIERS PARTNERSHIP FY13 MID-YEAR REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Burns, H.; Flach, G.; Langton, C.; KOSSON, D.; BROWN, K.; SAMSON, E.; MEEUSSEN, J.; SLOOT, H.; GARBOCZI, E.

    2013-05-01

    presentations from the CBP Software Toolbox Demonstration and User Workshop, which are briefly described below, can be accessed from the CBP webpage at http://cementbarriers.org/. The website was recently modified to describe the CBP Software Toolbox and includes an interest form for application to use the software. The CBP FY13 program is continuing research to improve and enhance the simulation tools as well as develop new tools that model other key degradation phenomena not addressed in Version 1.0. Also efforts to continue to verify the various simulation tools thru laboratory experiments and analysis of field specimens are ongoing to quantify and reduce the uncertainty associated with performance assessments are ongoing. This mid-year report also includes both a summary on the FY13 software accomplishments in addition to the release of Version 1.0 of the CBP Software Toolbox and the various experimental programs that are providing data for calibration and validation of the CBP developed software. The focus this year for experimental studies was to measure transport in cementitious material by utilization of a leaching method and reduction capacity of saltstone field samples. Results are being used to calibrate and validate the updated carbonation model.

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

    Directory of Open Access Journals (Sweden)

    Carmalin Sophia Ayyappan

    2015-06-01

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

  2. Mechanisms and modelling of waste-cement and cement-host rock interactions

    Science.gov (United States)

    2017-06-01

    Safe and sustainable disposal of hazardous and radioactive waste is a major concern in today's industrial societies. The hazardous waste forms originate from residues of thermal treatment of waste, fossil fuel combustion and ferrous/non-ferrous metal smelting being the most important ones in terms of waste production. Low- and intermediate-level radioactive waste is produced in the course of nuclear applications in research and energy production. For both waste forms encapsulation in alkaline, cement-based matrices is considered to ensure long-term safe disposal. Cementitious materials are in routine use as industrial materials and have mainly been studied with respect to their evolution over a typical service life of several decades. Use of these materials in waste management applications, however, requires assessments of their performance over much longer time periods on the order of thousands to several ten thousands of years.

  3. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-13

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lindle, Dennis W.; Shuh, David K.

    2005-02-28

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

  6. 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.

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

    Science.gov (United States)

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

    2006-05-01

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

  8. Experimental Study on Cementitious Composites Embedded with Organic Microcapsules

    Directory of Open Access Journals (Sweden)

    Zhiwei Qian

    2013-09-01

    Full Text Available The recovery behavior for strength and impermeability of cementitious composites embedded with organic microcapsules was investigated in this study. Mortar specimens were formed by mixing the organic microcapsules and a catalyst with cement and sand. The mechanical behaviors of flexural and compression strength were tested. The results showed that strength could increase by up to nine percent with the addition of a small amount of microcapsules and then decrease with an increasing amount of microcapsules. An orthogonal test for investigating the strength recovery rate was designed and implemented for bending and compression using the factors of water/cement ratio, amount of microcapsules, and preloading rate. It is shown that the amount of microcapsules plays a key role in the strength recovery rate. Chloride ion permeability tests were also carried out to investigate the recovery rate and healing effect. The initial damage was obtained by subjecting the specimens to compression. Both the recovery rate and the healing effect were nearly proportional to the amount of microcapsules. The obtained cementitious composites can be seen as self-healing owing to their recovery behavior for both strength and permeability.

  9. Experimental Study on Cementitious Composites Embedded with Organic Microcapsules.

    Science.gov (United States)

    Wang, Xianfeng; Xing, Feng; Zhang, Ming; Han, Ningxu; Qian, Zhiwei

    2013-09-16

    The recovery behavior for strength and impermeability of cementitious composites embedded with organic microcapsules was investigated in this study. Mortar specimens were formed by mixing the organic microcapsules and a catalyst with cement and sand. The mechanical behaviors of flexural and compression strength were tested. The results showed that strength could increase by up to nine percent with the addition of a small amount of microcapsules and then decrease with an increasing amount of microcapsules. An orthogonal test for investigating the strength recovery rate was designed and implemented for bending and compression using the factors of water/cement ratio, amount of microcapsules, and preloading rate. It is shown that the amount of microcapsules plays a key role in the strength recovery rate. Chloride ion permeability tests were also carried out to investigate the recovery rate and healing effect. The initial damage was obtained by subjecting the specimens to compression. Both the recovery rate and the healing effect were nearly proportional to the amount of microcapsules. The obtained cementitious composites can be seen as self-healing owing to their recovery behavior for both strength and permeability.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-01

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

  11. 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.

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, G.J.

    1975-08-05

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-09

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

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

    Energy Technology Data Exchange (ETDEWEB)

    SAMS TL; EDGE JA; SWANBERG DJ; ROBBINS RA

    2011-01-13

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

  17. Stabilization Using Phosphate Bonded Ceramics. Salt Containing Mixed Waste Treatment. Mixed Waste Focus Area. OST Reference No. 117

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1999-09-01

    Throughout the Department of Energy (DOE) complex there are large inventories of homogeneous mixed waste solids, such as wastewater treatment residues, fly ashes, and sludges that contain relatively high concentrations (greater than 15% by weight) of salts. The inherent solubility of salts (e.g., nitrates, chlorides, and sulfates) makes traditional treatment of these waste streams difficult, expensive, and challenging. One alternative is low-temperature stabilization by chemically bonded phosphate ceramics (CBPCs). The process involves reacting magnesium oxide with monopotassium phosphate with the salt waste to produce a dense monolith. The ceramic makes a strong environmental barrier, and the metals are converted to insoluble, low-leaching phosphate salts. The process has been tested on a variety of surrogates and actual mixed waste streams, including soils, wastewater, flyashes, and crushed debris. It has also been demonstrated at scales ranging from 5 to 55 gallons. In some applications, the CBPC technology provides higher waste loadings and a more durable salt waste form than the baseline method of cementitious grouting. Waste form test specimens were subjected to a variety of performance tests. Results of waste form performance testing concluded that CBPC forms made with salt wastes meet or exceed both RCRA and recommended Nuclear Regulatory Commission (NRC) low-level waste (LLW) disposal criteria. Application of a polymer coating to the CBPC may decrease the leaching of salt anions, but continued waste form evaluations are needed to fully assess the deteriorating effects of this leaching, if any, over time.

  18. 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

  19. 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.

  20. 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

  1. A review: Self-healing in cementitious materials and engineered cementitious composite as a self-healing material

    DEFF Research Database (Denmark)

    Wu, Min; Johannesson, Björn; Geiker, Mette

    2012-01-01

    Cementitious Composite (ECC) is a new type of cement-based materials, which has unique properties compared with traditional cementitious materials. Further, a summary about the research status of self-healing on ECC is given. It shows that ECC bears great potential in realizing effective self-healing due......Cementitious materials are the most widely used building materials all over the word. However, deterioration is inevitable even since the very beginning of the service life, then maintenance and repair work, which are often labor- and capital-intensive, would be followed. Thus, self-healing...... of the affected cementitious materials is of great importance. Self-healing phenomenon in cementitious materials has been noticed and been studying for a long time. The possible mechanisms for self-healing phenomenon in cementitious materials, which are summarized based on substantial experimental studies...

  2. Service life prediction and fibre reinforced cementitious composites

    DEFF Research Database (Denmark)

    Stoklund Larsen, E.

    The present Ph.D.thesis addresses the service life concept on the fibre reinforced cementitious composites. The advantages and problems of adding fibre to a cementitious matrix and the influence on service life are described. In SBI Report 221, Service life prediction and cementitious somposites......, the factors affecting the pure cementitious composite are described. Different sizes and types of fibre reinforced crmentitious composites have been chosen to illustrate different ageing and deterioration mechanisms. Some ageing mechanisms can be accelerated and others cannot which is demonstrated in a test...

  3. Leaching tests of simulated Cogema bituminized waste form

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    G. Becker; M. Connolly; M. McIlwain

    1999-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-28

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

  7. Can superabsorbent polymers mitigate shrinkage in cementitious materials blended with supplementary cementitious materials?

    DEFF Research Database (Denmark)

    Snoeck, Didier; Jensen, Ole Mejlhede; De Belie, Nele

    2016-01-01

    A promising way to mitigate autogenous shrinkage in cementitious materials with a low water-to-binder ratio is internal curing by the use of superabsorbent polymers. Superabsorbent polymers are able to absorb multiple times their weight in water and can be applied as an internal water reservoir...... shrinkage in materials blended with fly ash or blast-furnace slag remain scarce, especially after one week of age. This paper focuses on the autogenous shrinkage by performing manual and automated shrinkage measurements up to one month of age. Without superabsorbent polymers, autogenous shrinkage...... was reduced in cement pastes with the supplementary cementitious materials versus Portland cement pastes. At later ages, the rate of autogenous shrinkage is higher due to the pozzolanic activity of the supplementary cementitious materials. Internal curing by means of superabsorbent polymers is successful...

  8. 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

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-05-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-01

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

  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

    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).

  15. Innovative Structural Materials and Sections with Strain Hardening Cementitious Composites

    Science.gov (United States)

    Dey, Vikram

    The motivation of this work is based on development of new construction products with strain hardening cementitious composites (SHCC) geared towards sustainable residential applications. The proposed research has three main objectives: automation of existing manufacturing systems for SHCC laminates; multi-level characterization of mechanical properties of fiber, matrix, interface and composites phases using servo-hydraulic and digital image correlation techniques. Structural behavior of these systems were predicted using ductility based design procedures using classical laminate theory and structural mechanics. SHCC sections are made up of thin sections of matrix with Portland cement based binder and fine aggregates impregnating continuous one-dimensional fibers in individual or bundle form or two/three dimensional woven, bonded or knitted textiles. Traditional fiber reinforced concrete (FRC) use random dispersed chopped fibers in the matrix at a low volume fractions, typically 1-2% to avoid to avoid fiber agglomeration and balling. In conventional FRC, fracture localization occurs immediately after the first crack, resulting in only minor improvement in toughness and tensile strength. However in SHCC systems, distribution of cracking throughout the specimen is facilitated by the fiber bridging mechanism. Influence of material properties of yarn, composition, geometry and weave patterns of textile in the behavior of laminated SHCC skin composites were investigated. Contribution of the cementitious matrix in the early age and long-term performance of laminated composites was studied with supplementary cementitious materials such as fly ash, silica fume, and wollastonite. A closed form model with classical laminate theory and ply discount method, coupled with a damage evolution model was utilized to simulate the non-linear tensile response of these composite materials. A constitutive material model developed earlier in the group was utilized to characterize and

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-05-19

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

  17. 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.

  18. Hydration characteristics and environmental friendly performance of a cementitious material composed of calcium silicate slag.

    Science.gov (United States)

    Zhang, Na; Li, Hongxu; Zhao, Yazhao; Liu, Xiaoming

    2016-04-05

    Calcium silicate slag is an alkali leaching waste generated during the process of extracting Al2O3 from high-alumina fly ash. In this research, a cementitious material composed of calcium silicate slag was developed, and its mechanical and physical properties, hydration characteristics and environmental friendly performance were investigated. The results show that an optimal design for the cementitious material composed of calcium silicate slag was determined by the specimen CFSC7 containing 30% calcium silicate slag, 5% high-alumina fly ash, 24% blast furnace slag, 35% clinker and 6% FGD gypsum. This blended system yields excellent physical and mechanical properties, confirming the usefulness of CFSC7. The hydration products of CFSC7 are mostly amorphous C-A-S-H gel, rod-like ettringite and hexagonal-sheet Ca(OH)2 with small amount of zeolite-like minerals such as CaAl2Si2O8·4H2O and Na2Al2Si2O8·H2O. As the predominant hydration products, rod-like ettringite and amorphous C-A-S-H gel play a positive role in promoting densification of the paste structure, resulting in strength development of CFSC7 in the early hydration process. The leaching toxicity and radioactivity tests results indicate that the developed cementitious material composed of calcium silicate slag is environmentally acceptable. This study points out a promising direction for the proper utilization of calcium silicate slag in large quantities.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-13

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

  1. 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.

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-03-26

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

  4. Non-cementitious compositions comprising vaterite and methods thereof

    Energy Technology Data Exchange (ETDEWEB)

    Devenney, Martin; Fernandez, Miguel; Morgan, Samuel O.

    2015-09-15

    Non-cementitious compositions and products are provided. The compositions of the invention include a carbonate additive comprising vaterite such as reactive vaterite. Additional aspects of the invention include methods of making and using the non-cementitious compositions and products.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  6. Properties and Microstructure of Polymer Emulsions Modified Fibers Reinforced Cementitious Composites

    Institute of Scientific and Technical Information of China (English)

    WU Ying; SUN Qianyao; KONG Lian; FANG He

    2014-01-01

    The synthesis and characterization of a new class of cementitious composites filled with polymer emulsions were investigated, and their superior mechanical strength and durability properties compared to composites devoid of fillers were reported. Polymer emulsions were utilized to mechanically reinforce the composite and bridge the cement, fly ash, aggregate and fibers. The results reveal that the epoxy emulsion and poly (ethylene-co-vinyl acetate) emulsion markedly enhance the mechanical and durability properties of cemetitious composites. The fibers can be pulled out in the form of slip-hardening and the abrasion phenomenon can be observed clearly on the surface of the fibers. The hydration extent of cement is higher than that of the pristine composites. The polymer modified cementitious composites designed on micromechanics, have flexibility and plasticity which could be applied for a novel form of multifunctional materials with a range of pipeline coatings applications.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-04-15

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

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

    Science.gov (United States)

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

    2011-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  12. The Expanded Capabilities Of The Cementitious Barriers Partnership Software Toolbox Version 2.0 - 14331

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Heather; Flach, Greg; Smith, Frank; Langton, Christine; Brown, Kevin; Kosson, David; Samson, Eric; Mallick, Pramod

    2014-01-10

    The Cementitious Barriers Partnership (CBP) Project is a multi-disciplinary, multi-institutional collaboration supported by the U.S. Department of Energy (US DOE) Office of Tank Waste Management. The CBP program has developed a set of integrated tools (based on state-of-the-art models and leaching test methods) that help improve understanding and predictions of the long-term structural, hydraulic and chemical performance of cementitious barriers used in nuclear applications. The CBP Software Toolbox – “Version 1.0” was released early in FY2013 and was used to support DOE-EM performance assessments in evaluating various degradation mechanisms that included sulfate attack, carbonation and constituent leaching. The sulfate attack analysis predicted the extent and damage that sulfate ingress will have on concrete vaults over extended time (i.e., > 1000 years) and the carbonation analysis provided concrete degradation predictions from rebar corrosion. The new release “Version 2.0” includes upgraded carbonation software and a new software module to evaluate degradation due to chloride attack. Also included in the newer version are a dual regime module allowing evaluation of contaminant release in two regimes – both fractured and un-fractured. The integrated software package has also been upgraded with new plotting capabilities and many other features that increase the “user-friendliness” of the package. Experimental work has been generated to provide data to calibrate the models to improve the credibility of the analysis and reduce the uncertainty. Tools selected for and developed under this program have been used to evaluate and predict the behavior of cementitious barriers used in near-surface engineered waste disposal systems for periods of performance up to or longer than 100 years for operating facilities and longer than 1000 years for waste disposal. The CBP Software Toolbox is and will continue to produce tangible benefits to the working DOE

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

    Science.gov (United States)

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

    2014-05-01

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

  14. 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.

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

    Science.gov (United States)

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

    2011-12-01

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

  16. 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

  17. 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.

  18. Absorbency of Superabsorbent Polymers in Cementitious Environments

    DEFF Research Database (Denmark)

    Esteves, Luis Pedro; Jensen, Ole Mejlhede

    2012-01-01

    Optimal use of superabsorbent polymers (SAP) in cement-based materials relies on knowledge on how SAP absorbency is influenced by different physical and chemical parameters. These parameters include salt concentration in the pore fluid, temperature of the system and SAP particle size. The present...... composition of the exposure liquid is investigated with atomic absorption spectroscopy. The paper provides the reader with knowledge about the absorption capacity of SAP in a cementitious environment, and how the absorption process may influence the cement pore fluid....

  19. Chloride diffusion in partially saturated cementitious material

    DEFF Research Database (Denmark)

    Nielsen, Erik Pram; Geiker, Mette Rica

    2003-01-01

    The paper proposes a combined application of composite theory and Powers' model for microstructural development for the estimation of the diffusion coefficient as a function of the moisture content of a defect-free cementitious material. Measurements of chloride diffusion in mortar samples (440 kg....../m(3) rapid-hardening Portland cement, w/c = 0.5, maturity minimum 6 months) stored at 65% and 85% RH, as well as in vacuum-saturated mortar samples, illustrate the applicability of the method. (C) 2003 Elsevier Science Ltd. All rights reserved....

  20. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-03-27

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

  2. 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.

  3. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-01-17

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D.

    1996-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-03-31

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

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

    Science.gov (United States)

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

    2016-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D.

    1996-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  13. Final Report - Assessment of Potential Phosphate Ion-Cementitious Materials Interactions

    Energy Technology Data Exchange (ETDEWEB)

    Naus, Dan J [ORNL; Mattus, Catherine H [ORNL; Dole, Leslie Robert [ORNL

    2007-06-01

    The objectives of this limited study were to: (1) review the potential for degradation of cementitious materials due to exposure to high concentrations of phosphate ions; (2) provide an improved understanding of any significant factors that may lead to a requirement to establish exposure limits for concrete structures exposed to soils or ground waters containing high levels of phosphate ions; (3) recommend, as appropriate, whether a limitation on phosphate ion concentration in soils or ground water is required to avoid degradation of concrete structures; and (4) provide a "primer" on factors that can affect the durability of concrete materials and structures in nuclear power plants. An assessment of the potential effects of phosphate ions on cementitious materials was made through a review of the literature, contacts with concrete research personnel, and conduct of a "bench-scale" laboratory investigation. Results of these activities indicate that: no harmful interactions occur between phosphates and cementitious materials unless phosphates are present in the form of phosphoric acid; phosphates have been incorporated into concrete as set retarders, and phosphate cements have been used for infrastructure repair; no standards or guidelines exist pertaining to applications of reinforced concrete structures in high-phosphate environments; interactions of phosphate ions and cementitious materials has not been a concern of the research community; and laboratory results indicate similar performance of specimens cured in phosphate solutions and those cured in a calcium hydroxide solution after exposure periods of up to eighteen months. Relative to the "primer," a separate NUREG report has been prepared that provides a review of pertinent factors that can affect the durability of nuclear power plant reinforced concrete structures.

  14. 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

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

    Science.gov (United States)

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

    2007-02-01

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

  16. Strengthening masonry infill panels using engineered cementitious composites

    DEFF Research Database (Denmark)

    Dehghani, Ayoub; Fischer, Gregor; Nateghi Alahi, Fariborz

    2015-01-01

    This comprehensive experimental study aims at investigating the behavior of masonry infill panels strengthened by fiber reinforced engineered cementitious composites (ECC). The experimental program included testing of materials, masonry elements and panels. Material tests were carried out first f...

  17. Studies of ancient concrete as analogs of cementitious sealing materials for a repository in tuff

    Energy Technology Data Exchange (ETDEWEB)

    Roy, D.M.; Langton, C.A.

    1989-03-01

    The durability of ancient cementitious materials has been investigated to provide data applicable to determining the resistance to weathering of concrete materials for sealing a repository for storage of high-level radioactive waste. Because tuff and volcanic ash are used in the concretes in the vicinity of Rome, the results are especially applicable to a waste repository in tuff. Ancient mortars, plasters, and concretes collected from Rome, Ostia, and Cosa dating to the third century BC show remarkable durability. The aggregates used in the mortars, plasters, and concretes included basic volcanic and pyroclastic rocks (including tuff), terra-cotta, carbonates, sands, and volcanic ash. The matrices of ancient cementitious materials have been characterized and classified into four categories: (1) hydraulic hydrated lime and hydrated lime cements, (2) hydraulic aluminous and ferruginous hydrated lime cements ({plus_minus} siliceous components), (3) pozzolana/hydrated lime cements, and (4) gypsum cements. Most of the materials investigated are in category (3). The materials were characterized to elucidate aspects of the technology that produced them and their response to the environmental exposure throughout their centuries of existence. Their remarkable properties are the result of a combination of chemical, mineralogical, and microstructural factors. Their durability was found to be affected by the matrix mineralogy, particle size, and porosity; aggregate type, grading and proportioning; and the methodology of placement. 30 refs.

  18. Studies of ancient concrete as analogs of cementitious sealing materials for a repository in tuff

    Energy Technology Data Exchange (ETDEWEB)

    Roy, D.M.; Langton, C.A.

    1989-03-01

    The durability of ancient cementitious materials has been investigated to provide data applicable to determining the resistance to weathering of concrete materials for sealing a repository for storage of high-level radioactive waste. Because tuff and volcanic ash are used in the concretes in the vicinity of Rome, the results are especially applicable to a waste repository in tuff. Ancient mortars, plasters, and concretes collected from Rome, Ostia, and Cosa dating to the third century BC show remarkable durability. The aggregates used in the mortars, plasters, and concretes included basic volcanic and pyroclastic rocks (including tuff), terra-cotta, carbonates, sands, and volcanic ash. The matrices of ancient cementitious materials have been characterized and classified into four categories: (1) hydraulic hydrated lime and hydrated lime cements, (2) hydraulic aluminous and ferruginous hydrated lime cements ({plus_minus} siliceous components), (3) pozzolana/hydrated lime cements, and (4) gypsum cements. Most of the materials investigated are in category (3). The materials were characterized to elucidate aspects of the technology that produced them and their response to the environmental exposure throughout their centuries of existence. Their remarkable properties are the result of a combination of chemical, mineralogical, and microstructural factors. Their durability was found to be affected by the matrix mineralogy, particle size, and porosity; aggregate type, grading and proportioning; and the methodology of placement. 30 refs.

  19. Reactivity of nitrate and organic acids at the concrete–bitumen interface of a nuclear waste repository cell

    Energy Technology Data Exchange (ETDEWEB)

    Bertron, A., E-mail: bertron@insa-toulouse.fr [Université de Toulouse (France); UPS, INSA (France); LMDC (Laboratoire Matériaux et Durabilité des Constructions), 135, avenue de Rangueil, F-31 077, Toulouse Cedex 04 (France); Jacquemet, N. [Université de Toulouse (France); UPS, INSA (France); LMDC (Laboratoire Matériaux et Durabilité des Constructions), 135, avenue de Rangueil, F-31 077, Toulouse Cedex 04 (France); Erable, B. [Université de Toulouse (France); INPT, UPS (France); CNRS, Laboratoire de Génie Chimique, 4, Allée Emile Monso, F-31030 Toulouse (France); Sablayrolles, C. [Université de Toulouse (France); INP (France); LCA (Laboratoire de Chimie Agro-Industrielle), ENSIACET, 4 allée Emile Monso, BP 44 362, 31432 Toulouse Cedex 4 (France); INRA (France); LCA (Laboratoire de Chimie Agro-Industrielle), F-31029 Toulouse (France); Escadeillas, G. [Université de Toulouse (France); UPS, INSA (France); LMDC (Laboratoire Matériaux et Durabilité des Constructions), 135, avenue de Rangueil, F-31 077, Toulouse Cedex 04 (France); Albrecht, A. [Andra, 1-7, rue Jean-Monnet, 92298 Châtenay-Malabry (France)

    2014-03-01

    Highlights: • Interactions of cement paste and organic acid–nitrate solutions were investigated. • Cement leaching imposed alkaline pH (>10) very rapidly in the liquid media. • Acetic acid action on cement paste was similar to that of classical leaching. • Oxalic acid attack formed Ca-oxalate salts; organic matter in solution decreased. • Nitrate was stable under abiotic conditions and with organic matter. - Abstract: This study investigates the fate of nitrate and organic acids at the bitumen–concrete interface within repository cell for long-lived, intermediate-level, radioactive wastes. The interface was simulated by a multiphase system in which cementitious matrices (CEM V cement paste specimens) were exposed to bitumen model leachates consisting of nitrates and acetic acid with and without oxalic acid, chemical compounds likely to be released by bitumen. Leaching experiments were conducted with daily renewal of the solutions in order to accelerate reactions. The concentrations of anions (acetate, oxalate, nitrate, and nitrite) and cations (calcium, potassium) and the pH were monitored over time. Mineralogical changes of the cementitious matrices were analysed by XRD. The results confirmed the stability of nitrates in the abiotic conditions of the experiments. The action of acetic acid on the cementitious matrix was similar to that of ordinary leaching in the absence of organic acids (i.e. carried out with water or strong acids); no specific interaction was detected between acetate and cementitious cations. The reaction of oxalic acid with the cementitious phases led to the precipitation of calcium oxalate salts in the outer layer of the matrix. The concentration of oxalate was reduced by 65% inside the leaching medium.

  20. 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.

  1. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    BLACKFORD LT

    2008-02-04

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

  3. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-15

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

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

    Science.gov (United States)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-12-01

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

  7. 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.

  8. Wet-Treated MSWI Fly Ash Used as Supplementary Cementitious Material

    Directory of Open Access Journals (Sweden)

    Martin Keppert

    2015-01-01

    Full Text Available Municipal solid waste incineration (MSWI is a common technique in treatment of domestic waste. This technique annually produces approximately 25 Mt solid residues (i.e., bottom and fly ash worldwide which is also a major issue in current research. In this research we are concerned with reusing the fly ash (FA as supplementary cementitious material (SCM in concrete. Such application solves the problem with heavy metal immobilization as well. To remove the high content of undesired soluble salts, number of washing treatments has been applied. Chemical composition of FA has been examined before and after treatments. The impact of cement substitution by FA in concrete was evaluated by measurement of its compressive strength and durability.

  9. 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

  10. 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

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth J. Bateman; Charles W. Solbrig

    2006-07-01

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

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

    Science.gov (United States)

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

    2007-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-11-01

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

  15. New cementitious system: The case of glass frit

    Science.gov (United States)

    Fares, Galal

    Canada ranks as the world's third largest aluminium producer, and more than 80% of its aluminum industry is concentrated in Quebec. However, the spent pot-liner waste produced by the aluminium smelters accumulates with time into a considerable amount threatening the Canadian environment, especially that of Quebec. A new-engineered material, known as glass fit (GF) has been developed through the chemical treatment of such waste. GF shows potential hydraulic and pozzolanic properties. GF has been studied as a binder itself and as a supplementary cementitious material (SCM). The activation of industrial by-products into clinkerless binders is a novel trend that has attracted the attention of many researchers. The activation of GF into binder to produce paste, mortar and concrete was the first aim of this study. Potential activation of GF using different types and combinations of inorganic activators and temperatures of activation was successfully achieved and high strength concretes were obtained. Moreover, mortars with high compressive strength were obtained with well-formulated activators at ambient temperature. On the other hand, the utilization of industrial by-products as a partial replacement for cement in concrete is a widespread practice. As GF contains a high concentration of sodium in its structure, there is a concern as to the effect of sodium content on the development of alkali-silica reaction (ASR) expansion of concrete. Therefore, this study also aimed to investigate the effect of GF sodium content in the enhancement of ASR expansion and to find new synergistic mixtures that can effectively mitigate ASR expansion in the long term. We observed that ASR expansion decreases with the replacement level of GF. Different synergistic diagrams containing known SCM (silica fume, fly ash, and slag) were achieved from which different effective mixtures can effectively alleviate ASR expansion. In conclusion, the use of GF in the manufacture of concrete has great

  16. Temporary Cementitious Sealers in Enhanced Geothermal Systems

    Energy Technology Data Exchange (ETDEWEB)

    Sugama T.; Pyatina, T.; Butcher, T.; Brothers, L.; Bour, D.

    2011-12-31

    Unlike conventional hydrothennal geothermal technology that utilizes hot water as the energy conversion resources tapped from natural hydrothermal reservoir located at {approx}10 km below the ground surface, Enhanced Geothermal System (EGS) must create a hydrothermal reservoir in a hot rock stratum at temperatures {ge}200 C, present in {approx}5 km deep underground by employing hydraulic fracturing. This is the process of initiating and propagating a fracture as well as opening pre-existing fractures in a rock layer. In this operation, a considerable attention is paid to the pre-existing fractures and pressure-generated ones made in the underground foundation during drilling and logging. These fractures in terms of lost circulation zones often cause the wastage of a substantial amount of the circulated water-based drilling fluid or mud. Thus, such lost circulation zones must be plugged by sealing materials, so that the drilling operation can resume and continue. Next, one important consideration is the fact that the sealers must be disintegrated by highly pressured water to reopen the plugged fractures and to promote the propagation of reopened fractures. In response to this need, the objective of this phase I project in FYs 2009-2011 was to develop temporary cementitious fracture sealing materials possessing self-degradable properties generating when {ge} 200 C-heated scalers came in contact with water. At BNL, we formulated two types of non-Portland cementitious systems using inexpensive industrial by-products with pozzolanic properties, such as granulated blast-furnace slag from the steel industries, and fly ashes from coal-combustion power plants. These byproducts were activated by sodium silicate to initiate their pozzolanic reactions, and to create a cemetitious structure. One developed system was sodium silicate alkali-activated slag/Class C fly ash (AASC); the other was sodium silicate alkali-activated slag/Class F fly ash (AASF) as the binder of temper

  17. Mechanisms of cementitious material deterioration in biogas digester.

    Science.gov (United States)

    Voegel, C; Bertron, A; Erable, B

    2016-11-15

    Digesters produce biogas from organic wastes through anaerobic digestion processes. These digesters, often made of concrete, suffer severe premature deterioration caused mainly by the presence of fermentative microorganisms producing metabolites that are aggressive towards cementitious materials. To clarify the degradation mechanisms in an anaerobic digestion medium, ordinary Portland cement paste specimens were immersed in the liquid fraction of a running, lab-scale digester for 4weeks. The anaerobic digestion medium was a mixture of a biowaste substrate and sludge from municipal wastewater treatment plant used as a source of anaerobic bacteria. The chemical characteristics of the anaerobic digestion liquid phase were monitored over time using a pH metre, high performance liquid chromatography (HPLC) and ion chromatography (HPIC). An initial critical period of low pH in the bioreactors was observed before the pH stabilized around 8. Acetic, propionic and butyric acids were produced during the digestion with a maximum total organic acid concentration of 50mmolL(-1). The maximum ammonium content of the liquid phase was 40mmolL(-1), which was about seven times the upper limit of the highly aggressive chemical environment class (XA3) as defined by the European standard for the specification of concrete design in chemically aggressive environments (EN 206). The changes in the mineralogical, microstructural and chemical characteristics of the cement pastes exposed to the solid and liquid phase of the digesters were analysed at the end of the immersion period by X-ray diffraction (XRD), scanning electron microscopy (SEM) coupled with energy dispersive X-ray spectrometry (EDS) and electron-probe micro-analysis (EPMA). A 700-μm thick altered layer was identified in the cement paste specimens. The main biodeterioration patterns in the bioreactors' solid/liquid phase were calcium leaching and carbonation of the cement matrix.

  18. Bond strength of cementitious borehole plugs in welded tuff

    Energy Technology Data Exchange (ETDEWEB)

    Akgun, H.; Daemen, J.J.K. [Arizona Univ., Tucson, AZ (USA). Dept. of Mining and Geological Engineering

    1991-02-01

    Axial loads on plugs or seals in an underground repository due to gas, water pressures and temperature changes induced subsequent to waste and plug emplacement lead to shear stresses at the plug/rock contact. Therefore, the bond between the plug and rock is a critical element for the design and effectiveness of plugs in boreholes, shafts or tunnels. This study includes a systematic investigation of the bond strength of cementitious borehole plugs in welded tuff. Analytical and numerical analysis of borehole plug-rock stress transfer mechanics is performed. The interface strength and deformation are studied as a function of Young`s modulus ratio of plug and rock, plug length and rock cylinder outside-to-inside radius ratio. The tensile stresses in and near an axially loaded plug are analyzed. The frictional interface strength of an axially loaded borehole plug, the effect of axial stress and lateral external stress, and thermal effects are also analyzed. Implications for plug design are discussed. The main conclusion is a strong recommendation to design friction plugs in shafts, drifts, tunnels or boreholes with a minimum length to diameter ratio of four. Such a geometrical design will reduce tensile stresses in the plug and in the host rock to a level which should minimize the risk of long-term deterioration caused by excessive tensile stresses. Push-out tests have been used to determine the bond strength by applying an axial load to cement plugs emplaced in boreholes in welded tuff cylinders. A total of 130 push-out tests have been performed as a function of borehole size, plug length, temperature, and degree of saturation of the host tuff. The use of four different borehole radii enables evaluation of size effects. 119 refs., 42 figs., 20 tabs.

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

    Science.gov (United States)

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

    2017-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth J. Bateman; Charles W. Solbrig

    2008-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-04-01

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

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-02-01

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

  5. Study on cementitious properties of steel slag

    Directory of Open Access Journals (Sweden)

    Zhu G.

    2013-01-01

    Full Text Available The converter steel slag chemical and mineral components in China’s main steel plants have been analysed in the present paper. The electronic microscope, energy spectrum analysis, X-ray diffraction analysis confirmed the main mineral compositions in the converter slag. Converter slag of different components were grounded to obtain a powder with specific surface area over 400m2/kg, making them to take place some part of the cement in the concrete as the admixture and carry out the standard tests. The results indicate that the converter slag can be used as cementitious materials for construction. Furthermore, physical mechanic and durability tests on the concrete that certain amount of cement be substituted by converter steel slag powder from different steel plants are carried out, the results show that the concrete with partial substitution of steel slag powder has the advantages of higher later period strength, better frost resistance, good wear resistance and lower hydration heat, etc. This study can be used as the technical basis for “Steel Slag Powder Used For Cement And Concrete”, “Steel Slag Portland Cement”, “Low Heat Portland Steel Slag Cement”, “Steel Slag Road Cement” in China, as well as a driving force to the works of steel slag utilization with high-value addition, circular economy, energy conservation and discharge reduction in the iron and steel industry.

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-20

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

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

    Science.gov (United States)

    Mausolf, Edward John

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H.; Piepel, Gregory F.; Lindberg, Michael J.; Heasler, Patrick G.; Mercier, Theresa M.; Russell, Renee L.; Cozzi, Alex; Daniel, William E.; Eibling, Russell E.; Hansen, E. K.; Reigel, Marissa M.; Swanberg, David J.

    2013-09-30

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

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

    Science.gov (United States)

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

    2016-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-11

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

  19. Micro-structural characterization of the hydration products of bauxite-calcination-method red mud-coal gangue based cementitious materials.

    Science.gov (United States)

    Liu, Xiaoming; Zhang, Na; Yao, Yuan; Sun, Henghu; Feng, Huan

    2013-11-15

    In this research, the micro-structural characterization of the hydration products of red mud-coal gangue based cementitious materials has been investigated through SEM-EDS, (27)Al MAS NMR and (29)Si MAS NMR techniques, in which the used red mud was derived from the bauxite calcination method. The results show that the red mud-coal gangue based cementitious materials mainly form fibrous C-A-S-H gel, needle-shaped/rod-like AFt in the early hydration period. With increasing of the hydration period, densification of the pastes were promoted resulting in the development of strength. EDS analysis shows that with the Ca/Si of red mud-coal gangue based cementitious materials increases, the average Ca/Si and Ca/(Si+Al) atomic ratio of C-A-S-H gel increases, while the average Al/Si atomic ratio of C-A-S-H gel decreases. MAS NMR analysis reveals that Al in the hydration products of red mud-coal gangue based cementitious materials exists in the forms of Al(IV) and Al(VI), but mainly in the form of Al(VI). Increasing the Ca/Si ratio of raw material promotes the conversion of [AlO4] to [AlO6] and inhibits the combination between [AlO4] and [SiO4] to form C-A-S-H gel. Meanwhile, the polymerization degree of [SiO4] in the hydration products declines.

  20. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-09-23

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

  2. SCM Paste Samples Exposed To Aggressive Solutions. Cementitious Barriers Partnership

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-01

    This report summarizes experimental work performed by SIMCO Technologies Inc. (SIMCO) as part of the Cementitious Barriers Partnership (CBP) project. The test series followed an experimental program dedicated to the study of ordinary Portland cement (OPC) hydrated cement pastes exposed to aggressive solutions. In the present study, the scope is extended to hydrated cement pastes incorporating supplementary cementitious materials (SCM) such as fly ash and ground granulated blast furnace slag (GGBFS). Also, the range of aggressive contact solutions was expanded. The experimental program aimed at testing aggressive contact solutions that more closely mimic the chemical composition of saltstone pore solution. Five different solutions, some of which incorporated high levels of carbonate and nitrate, were placed in contact with four different hydrated cement paste mixes. In all solutions, 150 mmol/L of SO42– (14 400 ppm) were present. The solutions included different pH conditions and different sodium content. Two paste mixes were equivalent to Vault 1/4 and Vault 2 concrete mixes used at SRS in storage structures. Two additional paste mixes, cast at the same water-to-cement ratio and using the same cements but without SCMs, were also tested. The damage evolution in samples was monitored using ultrasonic pulse velocity (UPV) and mass measurements. After three and twelve months of exposure conditions, samples were taken out of solution containers and analyzed to perform migration tests and porosity measurements. Globally, results were in line with the previous study and confirmed that high pH may limit the formation of some deleterious phases like gypsum. In this case, ettringite may form but is not necessarily associated with damage. However, the high concentration of sodium may be associated with the formation of an AFm-like mineral called U-phase. The most significant evidences of damage were all associated with the Vault 2 paste analog. This

  3. SCM Paste Samples Exposed To Aggressive Solutions. Cementitious Barriers Partnership

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-01

    This report summarizes experimental work performed by SIMCO Technologies Inc. (SIMCO) as part of the Cementitious Barriers Partnership (CBP) project. The test series followed an experimental program dedicated to the study of ordinary Portland cement (OPC) hydrated cement pastes exposed to aggressive solutions. In the present study, the scope is extended to hydrated cement pastes incorporating supplementary cementitious materials (SCM) such as fly ash and ground granulated blast furnace slag (GGBFS). Also, the range of aggressive contact solutions was expanded. The experimental program aimed at testing aggressive contact solutions that more closely mimic the chemical composition of saltstone pore solution. Five different solutions, some of which incorporated high levels of carbonate and nitrate, were placed in contact with four different hydrated cement paste mixes. In all solutions, 150 mmol/L of SO42– (14 400 ppm) were present. The solutions included different pH conditions and different sodium content. Two paste mixes were equivalent to Vault 1/4 and Vault 2 concrete mixes uses at SRS in storage structures. Two additional paste mixes, cast at the same water-to-cement ratio and using the same cements but without SCMs, were also tested. The damage evolution in samples was monitored using ultrasonic pulse velocity (UPV) and mass measurements. After three and twelve months of exposure conditions, samples were taken out of solution containers and analyzed to perform migration tests and porosity measurements. Globally, results were in line with the previous study and confirmed that high pH may limit the formation of some deleterious phases like gypsum. In this case, ettringite may form but is not necessarily associated with damage. However, the high concentration of sodium may be associated with the formation of an AFm-like mineral called U-phase. The most significant evidences of damage were all associated with the Vault 2 paste analog. This

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-10

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-06

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

  6. Performance of alusilica as mineral admixture in cementitious systems

    DEFF Research Database (Denmark)

    Chi, Lin; Jensen, Ole Mejlhede

    2016-01-01

    This paper presents a preliminary study of the effect of alusilica (ALS) as a mineral admixture on the fresh properties and development of mechanical properties of cementitious systems. Cement was substituted with ALS with the ratio of 10% during grinding or blended during mixing. The produced ALS......, it is believed that ALS can be a useful cement substitution....

  7. Thermodynamics of Autogenous Self-healing in Cementitious Materials

    NARCIS (Netherlands)

    Huang, H.

    2014-01-01

    Concrete is a brittle composite cementitious material that easily fractures under tensile loading. Microcracks can appear throughout the concrete prior to application of any load because of temperature-induced strain and autogenous and drying shrinkage. There is no doubt that these cracks provide pr

  8. Decorative application of strain-hardening cementitious composites

    NARCIS (Netherlands)

    Ibrovic, V.; Lukovic, M.; Schlangen, E.

    2014-01-01

    Strain hardening cementitious composites (SHCC) have been used in variety of structural applications. Apart from this, they are also suitable for non-structural applications. In this work, the application of SHCC for producing cover plates for light switches and power sockets is presented. For desig

  9. Decorative application of strain-hardening cementitious composites

    NARCIS (Netherlands)

    Ibrovic, V.; Lukovic, M.; Schlangen, E.

    2014-01-01

    Strain hardening cementitious composites (SHCC) have been used in variety of structural applications. Apart from this, they are also suitable for non-structural applications. In this work, the application of SHCC for producing cover plates for light switches and power sockets is presented. For

  10. Thermodynamics of Autogenous Self-healing in Cementitious Materials

    NARCIS (Netherlands)

    Huang, H.

    2014-01-01

    Concrete is a brittle composite cementitious material that easily fractures under tensile loading. Microcracks can appear throughout the concrete prior to application of any load because of temperature-induced strain and autogenous and drying shrinkage. There is no doubt that these cracks provide

  11. Optimization of SO3 Content in Blended Cementitious Materials

    Institute of Scientific and Technical Information of China (English)

    CHEN Mei-zhu; LIANG Wen-quan; HE Zhen; LI Bei-xing

    2003-01-01

    Experimental investigation was conducted on the effects of gypsum types and SO3 content on the fluidity and strengths of different cementitious systems.The experimental results show that influences of gypsum in various cementitious materials are different.For cementitious materials blended with various proportions of slag-fly ash and 5% gypsum content,influences of gypsum and calcined gypsum on the fluidity and flexural/compressive strength are similar.It is revealed that "combination effect" and "synergistic effect" of slag and fly ash play an important role during hydration.For cementitious materials with 45% clinkers,30%slag,20%fly ash and 5%limestone,the optimized SO3 contents in gypsum and calcined gypsum are 3.13% and 3.51% respectively and the optimized gypsum content is 6.5%.While both of them are blended,the optimum ratio of gypsum to calcined gypsum is 40%∶60%(total gypsum content 6.5%),correspondingly the optimum ratio of SO3 is 19.3%∶32.4%.

  12. A Plastic Damage Mechanics Model for Engineered Cementitious Composites

    DEFF Research Database (Denmark)

    Dick-Nielsen, Lars; Stang, Henrik; Poulsen, Peter Noe

    2007-01-01

    This paper discusses the establishment of a plasticity-based damage mechanics model for Engineered Cementitious Composites (ECC). The present model differs from existing models by combining a matrix and fiber description in order to describe the behavior of the ECC material. The model provides in...

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hobbs, D.

    2012-02-24

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hobbs, D.

    2012-02-24

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

  16. Mechanical resilience and cementitious processes in Imperial Roman architectural mortar.

    Science.gov (United States)

    Jackson, Marie D; Landis, Eric N; Brune, Philip F; Vitti, Massimo; Chen, Heng; Li, Qinfei; Kunz, Martin; Wenk, Hans-Rudolf; Monteiro, Paulo J M; Ingraffea, Anthony R

    2014-12-30

    The pyroclastic aggregate concrete of Trajan's Markets (110 CE), now Museo Fori Imperiali in Rome, has absorbed energy from seismic ground shaking and long-term foundation settlement for nearly two millenia while remaining largely intact at the structural scale. The scientific basis of this exceptional service record is explored through computed tomography of fracture surfaces and synchroton X-ray microdiffraction analyses of a reproduction of the standardized hydrated lime-volcanic ash mortar that binds decimeter-sized tuff and brick aggregate in the conglomeratic concrete. The mortar reproduction gains fracture toughness over 180 d through progressive coalescence of calcium-aluminum-silicate-hydrate (C-A-S-H) cementing binder with Ca/(Si+Al) ≈ 0.8-0.9 and crystallization of strätlingite and siliceous hydrogarnet (katoite) at ≥ 90 d, after pozzolanic consumption of hydrated lime was complete. Platey strätlingite crystals toughen interfacial zones along scoria perimeters and impede macroscale propagation of crack segments. In the 1,900-y-old mortar, C-A-S-H has low Ca/(Si+Al) ≈ 0.45-0.75. Dense clusters of 2- to 30-µm strätlingite plates further reinforce interfacial zones, the weakest link of modern cement-based concrete, and the cementitious matrix. These crystals formed during long-term autogeneous reaction of dissolved calcite from lime and the alkali-rich scoriae groundmass, clay mineral (halloysite), and zeolite (phillipsite and chabazite) surface textures from the Pozzolane Rosse pyroclastic flow, erupted from the nearby Alban Hills volcano. The clast-supported conglomeratic fabric of the concrete presents further resistance to fracture propagation at the structural scale.

  17. Mechanical resilience and cementitious processes in Imperial Roman architectural mortar

    Science.gov (United States)

    Landis, Eric N.; Brune, Philip F.; Vitti, Massimo; Chen, Heng; Li, Qinfei; Kunz, Martin; Wenk, Hans-Rudolf; Monteiro, Paulo J. M.; Ingraffea, Anthony R.

    2014-01-01

    The pyroclastic aggregate concrete of Trajan’s Markets (110 CE), now Museo Fori Imperiali in Rome, has absorbed energy from seismic ground shaking and long-term foundation settlement for nearly two millenia while remaining largely intact at the structural scale. The scientific basis of this exceptional service record is explored through computed tomography of fracture surfaces and synchroton X-ray microdiffraction analyses of a reproduction of the standardized hydrated lime–volcanic ash mortar that binds decimeter-sized tuff and brick aggregate in the conglomeratic concrete. The mortar reproduction gains fracture toughness over 180 d through progressive coalescence of calcium–aluminum-silicate–hydrate (C-A-S-H) cementing binder with Ca/(Si+Al) ≈ 0.8–0.9 and crystallization of strätlingite and siliceous hydrogarnet (katoite) at ≥90 d, after pozzolanic consumption of hydrated lime was complete. Platey strätlingite crystals toughen interfacial zones along scoria perimeters and impede macroscale propagation of crack segments. In the 1,900-y-old mortar, C-A-S-H has low Ca/(Si+Al) ≈ 0.45–0.75. Dense clusters of 2- to 30-µm strätlingite plates further reinforce interfacial zones, the weakest link of modern cement-based concrete, and the cementitious matrix. These crystals formed during long-term autogeneous reaction of dissolved calcite from lime and the alkali-rich scoriae groundmass, clay mineral (halloysite), and zeolite (phillipsite and chabazite) surface textures from the Pozzolane Rosse pyroclastic flow, erupted from the nearby Alban Hills volcano. The clast-supported conglomeratic fabric of the concrete presents further resistance to fracture propagation at the structural scale. PMID:25512521

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-11

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-01

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

  1. Acceleration of percolation for cementitious sensors using conductive paint filler

    Science.gov (United States)

    Pinto, Irvin Jude Joseph

    Structural health monitoring has emerged as an important branch of civil engineering in recent times, with the need to automatically monitor structural performance over time to ensure structural integrity. More recently, the advent of smart sensing materials has given this field a major boost. Research has shown that smart sensing materials fabricated with conductive filler at a concentration close to the percolation threshold results in high sensitivity to strain due to the piezoresistive effect. Of particular interest to this research are cementitious sensors fabricated using carbon black fillers. Carbon black is considered because of its widespread availability and low cost over other conductive fillers such as carbon nanotubes and carbon nanofibers. A challenge in the fabrication of these sensors is that cementitious materials require a significant amount of carbon black to percolate, resulting in a loss in mechanical properties. This research investigates a new method to accelerate percolation of the materials, enabling cementitious sensors with fewer carbon black particles. A carbon black-based conductive paint that allows earlier percolation by facilitating conducting networks in cementitious sensors is used. The conductive paint consists of a block copolymer, SEBS (styrene-co-ethylene-co-butylene-co-styrene), filled with carbon black particles. The percolation thresholds of sensors fabricated both with and without conductive paint are, as well as their strain sensing characteristics and compressive strength. The study found that SEBS could successfully reduce the percolation threshold by 42%, and that samples with SEBS showed better electrical responses in dynamic conditions. Despite showing lower compressive strength, cementitious sensors fabricated with this novel conductive paint show promise for real time health monitoring applications.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, Aashish; Strachan, Denis M.

    2011-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-05-15

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

  5. Sequestration of phosphorus from wastewater by cement-based or alternative cementitious materials.

    Science.gov (United States)

    Wang, Xinjun; Chen, Jiding; Kong, Yaping; Shi, Xianming

    2014-10-01

    Cement-based and alternative cementitious materials were tested in the laboratory for their capability of removing phosphate from wastewater. The results demonstrated that both Langmuir and Freundlich adsorption isotherms were suitable for describing the adsorption characteristics of these materials. Among the four types of filter media tested, the cement-based mortar A has the highest value of maximum adsorption (30.96 mg g(-1)). The P-bonding energy (KL) and adsorption capacity (K) exhibited a positive correlation with the total content of Al2O3 and Fe2O3 in each mortar. The maximum amount of P adsorbed (Qm) and adsorption intensity (1/n) exhibited a positive correlation with the CaO content in each mortar. For three of them, the P-removal rates were in excess of 94 percent for phosphorus concentrations ranging from 20 to 1000 mg L(-1). The underlying mechanisms were examined using field emission scanning microscopy (FESEM), coupled with energy-dispersive X-ray spectroscopy (EDX) and X-ray powder diffraction (XRD). The results reveal that the removal of phosphate predominantly followed a precipitation mechanism in addition to weak physical interactions between the surface of adsorbent filter media and the metallic salts of phosphate. The use of cement-based or alternative cementitious materials in the form of ground powder shows great promise for developing a cost-effective and environmentally sustainable technology for P-sequestration and for wastewater treatment.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-01

    The U.S. Department of Energy’s Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford’s tank waste. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Supplemental Treatment is likely to be required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP’s LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750°C) continuous method by which LAW can be processed irrespective of whether the waste contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be comparable to LAW glass, i.e. leaches Tc-99, Re and Na at <2g/m2 during ASTM C1285 (Product Consistency) durability testing. Monolithing of the granular FBSR product was investigated to prevent dispersion during transport or burial/storage. Monolithing in an inorganic geopolymer binder, which is

  7. Monitoring early age cementitious materials using ultrasonic guided waves

    Science.gov (United States)

    Borgerson, Jacob L.

    The evaluation of early age concrete is critical for reducing construction times and ensuring quality. In this study, the use of ultrasonic guided waves for monitoring the development of early age cementitious materials is investigated. A torsional wave is transmitted and received through a waveguide that is embedded in early age mortar or concrete. As the cementitious material sets and hardens, the received wave(s) change, indicating the transition from a semifluid to a solid state. This thesis proposes two systems. The first system is a through-transmission system; a wave is transmitted on one end of an embedded waveguide using a sensor arrangement and then it is received on the opposite end of the rod with another sensor. This approach monitors the attenuation of the fundamental torsional wave mode, resulting from the leakage of energy from the cylindrical steel rod to the surrounding cementitious material. The evolution of the material's properties is related to the energy leakage or attenuation of the guided wave. The second system is a pulse-echo system; a wave is transmitted on one end of a partially embedded waveguide via a sensor arrangement that also receives the reflected signals. This approach monitors both the reflection from the end of the rod and the reflection from the point where the waveguide enters the material. The development of the cementitious material's mechanical properties is related to both the energy leaked into the surrounding material and the energy reflected at the point of entry. The ability of this method to only require access to one side of the specimen makes it attractive for monitoring early age cementitious materials in the field. Experiments were performed on mixtures with varying water-cement ratios (w/c = 0.40, 0.50, and 0.60), chemical admixtures (accelerant and retardant), mineral admixtures (silica fume and fly ash), and coarse aggregate (pea gravel). The time of setting and compressive strength of the various mixtures

  8. Immobilisation of Higher Activity Wastes from Nuclear Reactor Production of 99Mo

    Directory of Open Access Journals (Sweden)

    Martin W. A. Stewart

    2013-01-01

    Full Text Available A variety of intermediate- and low-level liquid and solid wastes are produced from reactor production of 99Mo using UAl alloy or UO2 targets and in principle can be collectively or individually converted into waste forms. At ANSTO, we have legacy acidic uranyl-nitrate-rich intermediate level waste (ILW from the latter, and an alkaline liquid ILW, a U-rich filter cake, plus a shorter lived liquid stream that rapidly decays to low-level waste (LLW standards, from the former. The options considered consist of cementitious products, glasses, glass-ceramics, or ceramics produced by vitrification or hot isostatic pressing for intermediate-level wastes. This paper discusses the progress in waste form development and processing to treat ANSTO’s ILW streams arising from 99Mo. The various waste forms and the reason for the process option chosen will be reviewed. We also address the concerns over adapting our chosen process for use in a hot-cell environment.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

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

    Science.gov (United States)

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

  12. Water absorption of superabsorbent polymers in a cementitious environment

    DEFF Research Database (Denmark)

    Jensen, Ole Mejlhede

    2011-01-01

    This paper focuses on the water absorption of superabsorbent polymers in a cementitious environment. The paper discusses different techniques to measure the water absorption capacity, and in particular it describes a technique which enables a simple and quick estimation of the water absorption...... capacity in a cementitious environment. The challenges met in defining the concept of water absorption capacity are treated, and the appropriateness of different types of superabsorbent polymers is also briefly dealt with. The concept “water absorption capacity” and its measurement seem straightforwardly...... simple, but a closer examination of the topic discloses many, significant difficulties. However, given proper cautiousness it is possible both to quickly estimate the water absorption capacity through a simple measurement as well as to examine how it will be influenced by different factors....

  13. Hybrid fiber reinforcement and crack formation in Cementitious Composite Materials

    DEFF Research Database (Denmark)

    Pereira, E.B.; Fischer, Gregor; Barros, J.A.O.

    2011-01-01

    reinforcement systems. The research described in this paper shows that the multi-scale conception of cracking and the use of hybrid fiber reinforcements do not necessarily result in an improved tensile behavior of the composite. Particular material design requirements may nevertheless justify the use of hybrid......The use of different types of fibers simultaneously for reinforcing cementitious matrices is motivated by the concept of a multi-scale nature of the crack propagation process. Fibers with different geometrical and mechanical properties are used to bridge cracks of different sizes from the micro......- to the macroscale. In this study, the performance of different fiber reinforced cementitious composites is assessed in terms of their tensile stress-crack opening behavior. The results obtained from this investigation allow a direct quantitative comparison of the behavior obtained from the different fiber...

  14. Progress in Research on Carbon Nanotubes Reinforced Cementitious Composites

    Directory of Open Access Journals (Sweden)

    Qinghua Li

    2015-01-01

    Full Text Available As one-dimensional (1D nanofiber, carbon nanotubes (CNTs have been widely used to improve the performance of nanocomposites due to their high strength, small dimensions, and remarkable physical properties. Progress in the field of CNTs presents a potential opportunity to enhance cementitious composites at the nanoscale. In this review, current research activities and key advances on multiwalled carbon nanotubes (MWCNTs reinforced cementitious composites are summarized, including the effect of MWCNTs on modulus of elasticity, porosity, fracture, and mechanical and microstructure properties of cement-based composites. The issues about the improvement mechanisms, MWCNTs dispersion methods, and the major factors affecting the mechanical properties of composites are discussed. In addition, large-scale production methods of MWCNTs and the effects of CNTs on environment and health are also summarized.

  15. Engineered cementitious composites for strengthening masonry infilled reinforced concrete frames

    DEFF Research Database (Denmark)

    Dehghani, Ayoub; Nateghi-Alahi, Fariborz; Fischer, Gregor

    2015-01-01

    The results of the second part of a comprehensive experimental program, aimed at investigating the behavior of masonry infilled reinforced concrete (RC) frames strengthened with fiber reinforced engineered cementitious composites (ECC) used as an overlay on the masonry wall, are presented in this...... and energy absorption capacity of the infilled frame, prevent brittle failure modes in the infill wall, and provide a reasonable system overstrength....

  16. Thermodynamics of Autogenous Self-healing in Cementitious Materials

    OpenAIRE

    Huang, H.

    2014-01-01

    Concrete is a brittle composite cementitious material that easily fractures under tensile loading. Microcracks can appear throughout the concrete prior to application of any load because of temperature-induced strain and autogenous and drying shrinkage. There is no doubt that these cracks provide preferential access for aggressive agents to penetrate into the concrete, probably causing corrosion of reinforcement steel and degradation of concrete. As a result, the service life of reinforced co...

  17. Characterization of supplementary cementitious materials by thermal analysis

    OpenAIRE

    Bernal, S. A.; Juenger, M.C.G.; Ke, X.; Matthes, W.; Lothenbach, B; De Belie, N.; Provis, J. L.

    2016-01-01

    Working Group 1 of RILEM TC 238-SCM ‘Hydration and microstructure of concrete with supplementary cementitious materials (SCMs)’ is defining best practices for the physical and chemical characterization of SCMs, and this paper focusses on their thermal analysis. Thermogravimetric analysis (TGA) can provide valuable data on the chemical and mineralogical composition of SCMs. Loss-on-ignition (LOI) testing is a commonly used, standardized, but less sophisticated version of TGA that measures mass...

  18. Development of an accurate pH measurement methodology for the pore fluids of low pH cementitious materials

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, M. C.; Garcia Calvo, J. L. [The Spanish National Research Council (CSIC), Madrid (Spain); Walker, C. [Japan Atomic Energy Agency (JAEA), Ibaraki (Japan)] [and others

    2012-08-15

    The main objective of this project has been the development of an agreed set of protocols for the pH measurement of the pore fluid of a low pH cementitious material. Three protocols have been developed (Chapter 2), a reference method, based on pore fluid expression (PFE), and two routine methods with and without filtering, based on Ex Situ Leaching (ESL) procedures. Templates have been designed on which to record details of the pH measurement for the reference (PFE) method (Appendix C) and the routine (ESL) methods without and with filtering (Appendix D). Preliminary protocols were based on a broad review of the literature (Appendix A) and refined through a series of test experiments of the more critical parameters (Appendix B). After definition of the preliminary protocols, two phases of interlaboratory tests were performed. The first phase (Chapter 3) used the same low pH cement paste and enabled the nine participating laboratories to use, become familiar with and to identify any problems/uncertainties in the preliminary protocols. The reported pH values were subjected to a statistical analysis of the (within laboratory) repeatability and (between-laboratory) reproducibility and so provided a reliability test of the preliminary protocols. The second phase (Chapter 4) of interlaboratory tests used four different candidate low pH cementitious materials in the same nine laboratories, which allowed testing, validation and comparison of the reported pH values, which were obtained using the final protocols for the reference (PFE) and routine (ESL) methods by statistical analysis. The proposed final protocols (Chapter 2) have resulted in the reported pH values having low deviation and high reproducibility and repeatability. This will allow confidence in the pH value when selecting a candidate low pH cementitious material to be used in the engineered component of a high-level nuclear waste repository.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-08-12

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

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

    Science.gov (United States)

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

    2010-06-01

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

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

    Science.gov (United States)

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

    2016-11-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2003-06-12

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

  3. Mechanical properties of gangue-containing aluminosilicate based cementitious materials

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    High performance aluminosilicate based cementitious materials were produced using calcined gangue as one of the major raw materials.The gangue was calcined at 500℃.The main constituent was calcined gangue, fly ash and slag, while alkali-silicate solutions were used as the diagenetic agent.The structure of gangue-containing aluminosilicate based cementitious materials was studied by the methods of IR, NMR and SEM.The results show that the mechanical properties are affected by the mass ratio between the gangue, slag and fly ash, the kind of activator and additional salt.For 28-day curing time, the compressive strength of the sample with a mass proportion of 2:1:1 (gangue: slag: fly ash) is 58.9 MPa, while the compressive strength of the sample containing 80wt%gangue can still be up to 52.3 MPa.The larger K+ favors the formation of large silicate oligomers with which Al(OH)4- prefers to bind.Therefore, in Na-K compounding activator solutions more oligomers exist which result in a stronger compressive strength of aluminosilicate-based cementitious materials than in the case of Na-containing activator.The reasons for this were found through IR and NMR analysis.Glauber's salt reduces the 3-day compressive strength of the paste, but increases its 7-day and 28-day compressive strengths.

  4. Dense packing properties of mineral admixtures in cementitious material

    Institute of Scientific and Technical Information of China (English)

    Yanzhou Peng; Shuguang Hu; Qingjun Ding

    2009-01-01

    The effect of ultra-fine fly ash (UFFA), steel slag (SS) and silica fume (SF) on packing density of binary, ternary and quaternary cementitious materials was studied in this paper in terms of minimum water requirement of cement. The influence of mineral admixtures on the relative density of pastes with low water/binder ratios was analyzed and the relationship between paste density and compressive strength of the corresponding hardened mortars was discussed. The results indicate that the incorporation of mineral admixtures can effectively improve the packing density of cementitious materials; the increase in packing density of a composite with incorporation of two or three kinds of mineral admixtures is even more obvious than that with only one mineral admixture. Moreover, an optimal amount of mineral admixture imparts to the mixture maximum packing density. The dense packing effect of a mineral admixture can increase the packing density of the resulting cementitious material and also the density of paste with low water/binder ratio, which evidently enhances the compressive strength of the hardened mortar.

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

    Directory of Open Access Journals (Sweden)

    Šogorov Stevan

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-03-19

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Harvey, K.B.

    1995-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-01

    The U.S. Department of Energy’s Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford’s tank waste. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Supplemental Treatment is likely to be required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP’s LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750°C) continuous method by which LAW can be processed irrespective of whether the waste contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be comparable to LAW glass, i.e. leaches Tc-99, Re and Na at <2g/m2 during ASTM C1285 (Product Consistency) durability testing. Monolithing of the granular FBSR product was investigated to prevent dispersion during transport or burial/storage. Monolithing in an inorganic geopolymer binder, which is

  9. Decommissioning and safety issues of liquid-mercury waste generated from high power spallation sources with particle accelerators

    CERN Document Server

    Chiriki, S; Odoj, R; Moormann, R; Hinssen, H. K; Bukaemskiy, A

    2009-01-01

    Large spallation sources are intended to be constructed in Europe (EURISOL nuclear physics facility and ESS-European Spallation Source). These facilities accumulate more than 20 metric tons of irradiated mercury in the target, which has to be treated as highly radioactive and chemo-toxic waste. Because solids are the only appropriate (immobile) form for this radiotoxic and toxic type of waste solidification is required for irradiated mercury. Our irradiation experimental studies on mercury waste revealed that mercury sulfide is a reasonable solid for disposal and shows larger stability in assumed accidents with water ingress in a repository compared to amalgams. For preparation of mercury sulfide a wet process is more suitable than a dry one. It is easier to perform under hot cell conditions and allows complete Hg-conversion. Embedding HgS in a cementitious matrix increases its stability.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-30

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

  11. Long-term effect on the solidified degraded cellulose-based waste slurry in cement matrix

    Directory of Open Access Journals (Sweden)

    Hosam El-Din Mostafa Saleh

    2009-03-01

    Full Text Available The long-term effects on solidification/stabilization (s/s of the secondary wastes, resulting from the oxidative degradation of some solid cellulosic-based wastes, in Portland Cement (CEM I have been investigated by X-ray diffraction (X-RD and Fourier transform infrared spectroscopy (FT-IR techniques. The effect of seven years leaching of the cemented waste forms obtained was carried out to assess the long-term immobilization behavior of the radionuclide in the solidified/stabilized waste that maybe exposed to fresh, ground or sea water.The results of this study confirm our previously published work that the oxidative degradation treatment of some cellulosic-based wastes is essential before incorporating into the cementitious inert matrix. In addition, the release of radionuclides from the cemented waste form is a diffusion controlling process, after the first washing out period lasting for nearly thirty days.Based on the results so far obtained it is concluded that Portland Cement could be considered as a potential inert matrix to immobilize the degraded cellulosic-based wastes for a short or long time of storage or a final disposal.

  12. Glass cullet as a new supplementary cementitious material (SCM)

    Science.gov (United States)

    Mirzahosseini, Mohammadreza

    Finely ground glass has the potential for pozzolanic reactivity and can serve as a supplementary cementitious material (SCM). Glass reaction kinetics depends on both temperature and glass composition. Uniform composition, amorphous nature, and high silica content of glass make ground glass an ideal material for studying the effects of glass type and particle size on reactivity at different temperature. This study focuses on how three narrow size ranges of clear and green glass cullet, 63--75 mum, 25--38 mum, and smaller than 25 mum, as well as combination of glass types and particle sizes affects the microstructure and performance properties of cementitious systems containing glass cullet as a SCM. Isothermal calorimetry, chemical shrinkage, thermogravimetric analysis (TGA), quantitative analysis of X-ray diffraction (XRD), and analysis of scanning electron microscope (SEM) images in backscattered (BS) mode were used to quantify the cement reaction kinetics and microstructure. Additionally, compressive strength and water sorptivity experiments were performed on mortar samples to correlate reactivity of cementitious materials containing glass to the performance of cementitious mixtures. A recently-developed modeling platform called "muic the model" was used to simulated pozzolanic reactivity of single type and fraction size and combined types and particle sizes of finely ground glass. Results showed that ground glass exhibits pozzolanic properties, especially when particles of clear and green glass below 25 mum and their combination were used at elevated temperatures, reflecting that glass cullet is a temperature-sensitive SCM. Moreover, glass composition was seen to have a large impact on reactivity. In this study, green glass showed higher reactivity than clear glass. Results also revealed that the simultaneous effect of sizes and types of glass cullet (surface area) on the degree of hydration of glass particles can be accounted for through a linear addition

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-06-27

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, P; Roberts, S; Bourcier, W

    2000-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

  19. Image-based detection and analysis of crack propagation in cementitious composites

    DEFF Research Database (Denmark)

    Pereira, E.B.; Fischer, Gregor; Barros, J.A.O.

    2011-01-01

    after the cracking process has occurred. The alternative nondestructive methods are often either not precise enough or experimentally too demanding. In this study, the use of an image analysis procedure to capture the crack initiation and propagation process is described, which utilizes digital images......The initiation and propagation of cracking in concrete and other cementitious materials is a governing mechanism for many physical and mechanical material properties. The observation of these cracking processes in concrete is typically taking place at discrete locations using destructive methods...... of the concrete while undergoing the cracking process. The results obtained with this method have shown that it is possible to monitor relatively small displacements on the specimen surface independently of the scale of the representative area of interest. The formed cracks are visible at relatively small crack...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

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

    Science.gov (United States)

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

    2004-02-01

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

  2. SECONDARY WASTE/ETF (EFFLUENT TREATMENT FACILITY) PRELIMINARY PRE-CONCEPTUAL ENGINEERING STUDY

    Energy Technology Data Exchange (ETDEWEB)

    MAY TH; GEHNER PD; STEGEN GARY; HYMAS JAY; PAJUNEN AL; SEXTON RICH; RAMSEY AMY

    2009-12-28

    This pre-conceptual engineering study is intended to assist in supporting the critical decision (CD) 0 milestone by providing a basis for the justification of mission need (JMN) for the handling and disposal of liquid effluents. The ETF baseline strategy, to accommodate (WTP) requirements, calls for a solidification treatment unit (STU) to be added to the ETF to provide the needed additional processing capability. This STU is to process the ETF evaporator concentrate into a cement-based waste form. The cementitious waste will be cast into blocks for curing, storage, and disposal. Tis pre-conceptual engineering study explores this baseline strategy, in addition to other potential alternatives, for meeting the ETF future mission needs. Within each reviewed case study, a technical and facility description is outlined, along with a preliminary cost analysis and the associated risks and benefits.

  3. Micro-structural characterization of the hydration products of bauxite-calcination-method red mud-coal gangue based cementitious materials

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiaoming [State Key Laboratory of Advanced Metallurgy, School of Metallurgical and Ecological Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Zhang, Na [Green Construction Materials and Circulation Economy Center, Architectural Design and Research Institute of Tsinghua University Co., Ltd., Beijing 100084 (China); Yao, Yuan, E-mail: yuanyaocas@163.com [School of Engineering and Computer Science, University of the Pacific, Stockton, CA 95211 (United States); Sun, Henghu; Feng, Huan [School of Engineering and Computer Science, University of the Pacific, Stockton, CA 95211 (United States)

    2013-11-15

    Highlights: • Al{sup IV} and Al{sup VI} both exist in the hydration products. • Increase of Ca/Si ratio promotes the conversion from [AlO{sub 4}] to [AlO{sub 6}]. • Polymerization degree of [SiO{sub 4}] in the hydration products declines. -- Abstract: In this research, the micro-structural characterization of the hydration products of red mud-coal gangue based cementitious materials has been investigated through SEM-EDS, {sup 27}Al MAS NMR and {sup 29}Si MAS NMR techniques, in which the used red mud was derived from the bauxite calcination method. The results show that the red mud-coal gangue based cementitious materials mainly form fibrous C-A-S-H gel, needle-shaped/rod-like AFt in the early hydration period. With increasing of the hydration period, densification of the pastes were promoted resulting in the development of strength. EDS analysis shows that with the Ca/Si of red mud-coal gangue based cementitious materials increases, the average Ca/Si and Ca/(Si + Al) atomic ratio of C-A-S-H gel increases, while the average Al/Si atomic ratio of C-A-S-H gel decreases. MAS NMR analysis reveals that Al in the hydration products of red mud-coal gangue based cementitious materials exists in the forms of Al{sup IV} and Al{sup VI}, but mainly in the form of Al{sup VI}. Increasing the Ca/Si ratio of raw material promotes the conversion of [AlO{sub 4}] to [AlO{sub 6}] and inhibits the combination between [AlO{sub 4}] and [SiO{sub 4}] to form C-A-S-H gel. Meanwhile, the polymerization degree of [SiO{sub 4}] in the hydration products declines.

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

    Energy Technology Data Exchange (ETDEWEB)

    Liu, G.

    1997-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-02-26

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

  6. Hybrid fiber reinforcement and crack formation in Cementitious Composite Materials

    DEFF Research Database (Denmark)

    Pereira, E.B.; Fischer, Gregor; Barros, J.A.O.

    2011-01-01

    reinforcement systems. The research described in this paper shows that the multi-scale conception of cracking and the use of hybrid fiber reinforcements do not necessarily result in an improved tensile behavior of the composite. Particular material design requirements may nevertheless justify the use of hybrid......- to the macroscale. In this study, the performance of different fiber reinforced cementitious composites is assessed in terms of their tensile stress-crack opening behavior. The results obtained from this investigation allow a direct quantitative comparison of the behavior obtained from the different fiber...

  7. Shear behavior of reinforced Engineered Cementitious Composites (ECC) beams

    DEFF Research Database (Denmark)

    Paegle, Ieva; Fischer, Gregor

    2010-01-01

    This paper describes an experimental investigation of the shear behavior of beams consisting of steel reinforced Engineered Cementitious Composites (ECC). Based on the strain hardening and multiple cracking behavior of ECC, this study investigates the extent to which ECC can improve the shear...... randomly distributed PVA (polyvinyl alcohol) fiber beams with different stirrup spacing and reinforced concrete (RC) beams for comparison. Displacement and strain measurements taken using the ARAMIS photogrammetric data acquisition system by means of processing at high frame rate captured images of applied...

  8. Crushing damage estimation for pavement with lightly cementitious bases

    CSIR Research Space (South Africa)

    De Beer, Morris

    2014-07-01

    Full Text Available , Pretoria, South Africa. De Beer, M. 2013. Revision of the South African Pavement Design Method. Project Focus Area: Bound Materials: Damage Laws. Cementitious Materials: Expanded/Revised Damage Laws for Crushing Failure. Contract Report SANRAL/SAPDM-D3... stream_source_info De Beer_2014.pdf.txt stream_content_type text/plain stream_size 25820 Content-Encoding UTF-8 stream_name De Beer_2014.pdf.txt Content-Type text/plain; charset=UTF-8 CRUSHING DAMAGE ESTIMATION...

  9. Talc-based cementitious products: Effect of talc calcination

    OpenAIRE

    2015-01-01

    This study reports the use of calcined talc for cementitious products making. The calcination is used to enhance the availability of magnesium from talc to react with phosphate for cement phase formation. It is shown that previous calcination of talc leads to products having enhanced mechanical performance due to the formation of more cement phase than in products based on raw talc. Talc fired at 900 °C was found to be the one in which magnesium release was maximal. Firing at temperature high...

  10. Technetium Sorption by Cementitious Materials Under Reducing Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Daniel I. [Savannah River Site (SRS), Aiken, SC (United States); Estes, Shanna L. [Environmental Engineering and Earth Sciences, Clemson University, Clemson, SC (United States); Powell, Brian A. [Environmental Engineering and Earth Sciences, Clemson University, Clemson, SC (United States)

    2012-09-28

    The objective of this study was to measure technetium ({sup 99}Tc) sorption to cementitious materials under reducing conditions to simulate Saltstone Disposal Facility conditions. {sup 99}Tc(VII) batch sorption experiments were conducted for 319 days in an inert glovebag with a variety of cementitious materials (aged cement, Vault 2, TR545, and TR547) containing varying amounts of blast furnace slag. Between 154 and 319 days, the {sup 99}Tc aqueous concentrations tended to remain constant and samples amended with different initial {sup 99}Tc concentrations, tended to merge at about 10{sup -9} M for Vault 2 (17% slag) and TR545 (90% slag) and 10{sup -8} M for TR547 (45% slag). This data provided strong evidence that solubility, and not adsorption (K{sub d} values), was controlling aqueous {sup 99}Tc concentrations. Laboratory data superimposed over thermodynamic speciation diagrams further supported the conclusion that solubility, and not adsorption controlled {sup 99}Tc aqueous concentrations. The oxidation state of the aqueous {sup 99}Tc at the end of the sorption experiment was determined by solvent extraction to be almost entirely {sup 99}Tc(VII). The pH of the present system was ~11.8. Previously proposed solubility controlling phases including Tc-sulfides may be present, but do not appear to control solubility. After the 319 day sorption period, the suspensions were removed from the glovebag and a desorption step under oxic conditions was conducted for 20 days by adding oxic, pH-buffered solutions to the suspensions. {sup 99}Tc aqueous concentrations increased by more than an order of magnitude and Eh increased by several hundred millivolts within 24 hours after the introduction of the oxic solutions. These desorption results are consistent with re-oxidation and dissolution/desorption of {sup 99}Tc(IV) phases possibly present in the cementitious materials after the anoxic sorption step of the experiment. Aqueous {sup 99}Tc concentrations continued to increase

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-06-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-03-31

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

  13. Designing added functions in engineered cementitious composites

    Science.gov (United States)

    Yang, En-Hua

    In this dissertation, a new and systematic material design approach is developed for ECC with added functions through material microstructures linkage to composite macroscopic behavior. The thesis research embodies theoretical development by building on previous ECC micromechanical models, and experimental investigations into three specific new versions of ECC with added functions aimed at addressing societal demands of our built infrastructure. Specifically, the theoretical study includes three important ECC modeling elements: Steady-state crack propagation analyses and simulation, predictive accuracy of the fiber bridging constitutive model, and development of the rate-dependent strain-hardening criteria. The first element establishes the steady-state cracking criterion as a fundamental requirement for multiple cracking behavior in brittle matrix composites. The second element improves the accuracy of crack-width prediction in ECC. The third element establishes the micromechanics basis for impact-resistant ECC design. Three new ECCs with added functions were developed and experimentally verified in this thesis research through the enhanced theoretical framework. A green ECC incorporating a large volume of industrial waste was demonstrated to possess reduced crack width and drying shrinkage. The self-healing ECC designed with tight crack width was demonstrated to recover transport and mechanical properties after microcrack damage when exposed to wet and dry cycles. The impact-resistant ECC was demonstrated to retain tensile ductility with increased strength under moderately high strain-rate loading. These new versions of ECC with added functions are expected to contribute greatly to enhancing the sustainability, durability, and safety of civil infrastructure built with ECC. This research establishes the effectiveness of micromechanics-based design and material ingredient tailoring for ECC with added new attributes but without losing its basic tensile ductile

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

    Science.gov (United States)

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

  15. Dynamic damage and stress-strain relations of ultra-high performance cementitious composites subjected to repeated impact

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    Ultra-high performance cementitious composites (UHPCC) were prepared by replacing 60% of cement with ultra-fine industrial waste powders.The dynamic damage and compressive stress-strain relations of UHPCC were studied using split Hopkinson pressure bar (SHPB).The damage of UHPCC subjected to repeated impact was measured by the ultrasonic pulse velocity method.Results show that the dynamic damage of UHPCC increases linearly with impact times and the abilities of repeated impact resistance of UHPCC are improved with increasing fiber volume fraction.The stress waves on impact were recorded and the average stress,strain and strain rate of UHPCC were calculated based on the wave propagation theory.The effects of strain rate,fibers volume fraction and impact times on the stress-strain relations of UHPCC were studied.Results show that the peak stress and elastic modulus decrease while the strain rate and peak strain increase gradually with increasing impact times.

  16. Effect of silicate solutions on metakaolinite based cementitious material

    Institute of Scientific and Technical Information of China (English)

    XIAO Xue-jun; LI Hua-jian; SUN Heng-hu

    2006-01-01

    High performance metakaolinite based cementitious materials were prepared with metakaolinite as main component, and the different modules of Na and Na-K silicate solutions as diagenetic agent. The results show that the mechanical properties are affected by different silicate solutions, compressive strengths of pastes hydrated for 3 d and 28 d with Na-K silicate solution (The modulus is 1) are about 43.68 and 78.52 MPa respectively. By analyzing the mechanical properties of Metakaolinite based cementitious materials, the diagenetic effect of lower module is better than higher module, and Na-K silicate solution is better than Na silicate solution. The structure of the Na and Na-K silicate solutions is studied with IR and 29Si NMR, the reason of the lower module and Na-K silicate solution improving the mechanical properties is that the low module silicate solution has lower polymeric degree of silicon dioxide, and the higher polymeric degree of silicon oxide tetrahedron(Q4) in Na-K silicate solution is less than Na silicate solution.

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

    Science.gov (United States)

    Nam, Sangchul; Namkoong, Wan

    2012-01-15

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

  18. Preparation of New Cementitious System using Fly Ash and Dehydrated Autoclaved Aerated Concrete

    Institute of Scientific and Technical Information of China (English)

    SHUI Zhonghe; LU Jianxin; TIAN Sufang; SHEN Peiliang; DING Sha

    2014-01-01

    We experimentally studied the interaction between pozzolanic material (fly ash) and dehydrated autoclaved aerated concrete (DAAC). The DAAC powder was obtained by grinding aerated concrete waste to particles finer than 75μm and was then heated to temperatures up to 900℃. New cementitious material was prepared by proportioning fly ash and DAAC, named as AF. X-ray diffraction (XRD) was employed to identify the crystalline phases of DAAC before and after rehydration. The hydration process of AF was analyzed by the heat of hydration and non-evaporable water content (Wn). The experimental results show that the highest reactivity of DAAC can be obtained by calcining the powder at 700℃and the dehydrated products are mainlyβ-C2S and CaO. The cumulative heat of hydration and Wn was found to be strongly dependent on the replacement level of fly ash, increasing the replacement level of fly ash lowered them in AF. The strength contribution rates on pozzolanic effect of fly ash in AF are always negative, showing a contrary tendency of that of cement-fly ash system.

  19. Technetium Sorption By Cementitious Materials Under Reducing Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Daniel I. [Savannah River National Lab., Aiken, SC (United States); Estes, Shanna L. [Clemson Univ., SC (United States). Environmental Engineering and Earth Sciences; Arai, Yuji [Clemson Univ., SC (United States). College of Agriculture, Forestry and Life Sciences; Powell, Brian A. [Clemson Univ., SC (United States). Environmental Engineering and Earth Sciences

    2013-07-18

    The objective of this study was to measure Tc sorption to cementitious materials under reducing conditions to simulate Saltstone Disposal Facility conditions. Earlier studies were conducted and the experimental conditions were found not to simulate those of the facility. Through a five month subcontract with Clemson University, sorption of {sup 99}Tc to four cementitious materials was examined within an anaerobic glovebag targeting a 0.1% H{sub 2}(g)/ 99.9% N{sub 2}(g) atmosphere. Early experiments based on Tc sorption and Eh indicated that 0.1% H{sub 2}(g) (a reductant) was necessary to preclude experimental impacts from O{sub 2}(g) diffusion into the glovebag. Preliminary data to date (up to 56 days) indicates that sorption of {sup 99}Tc to cementitious materials increased with increasing slag content for simulated saltstone samples. This is consistent with the conceptual model that redox active sulfide groups within the reducing slag facilitate reduction of Tc(VII) to Tc(IV). These experiments differ from previous experiments where a 2% H{sub 2}(g) atmosphere was maintained (Kaplan et al., 2011 (SRNL-STI-2010-00668)). The impact of the 2% H{sub 2}(g) reducing atmosphere on this data was examined and determined to cause the reduction of Tc in experimental samples without slag. In the present ongoing study, after 56 days, Tc sorption by the 50-year old cement samples (no slag) was undetectable, whereas Tc sorption in the cementitious materials containing slag continues to increase with contact time (measured after 1, 4, 8, 19 and 56 days). Sorption was not consistent with spike concentrations and steady state has not been demonstrated after 56 days. The average conditional K{sub d} value for the Vault 2 cementitious material was 873 mL/g (17% slag), for the TR547 Saltstone (45% slag) the conditional K{sub d} was 168 mL/g, and for TR545 (90% slag) the conditional K{sub d} was 1,619 mL/g. It is anticipated that additional samples will be collected until steady state

  20. Development of Ecoefficient Engineered Cementitious Composites Using Supplementary Cementitious Materials as a Binder and Bottom Ash Aggregate as Fine Aggregate

    Directory of Open Access Journals (Sweden)

    Jin Wook Bang

    2015-01-01

    Full Text Available The purpose of this study is to develop ecoefficient engineered cementitious composites (ECC using supplementary cementitious materials (SCMs, including fly ash (FA and blast furnace slag (SL as a binder material. The cement content of the ECC mixtures was replaced by FA and SL with a replacement rate of 25%. In addition, the fine aggregate of the ECC was replaced by bottom ash aggregate (BA with a substitution rate of 10%, 20%, and 30%. The influences of ecofriendly aggregates on fresh concrete properties and on mechanical properties were experimentally investigated. The test results revealed that the substitution of SCMs has an advantageous effect on fresh concrete’s properties; however, the increased water absorption and the irregular shape of the BA can potentially affect the fresh concrete’s properties. The substitution of FA and SL in ECC led to an increase in frictional bond at the interface between PVA fibers and matrix, improved the fiber dispersion, and showed a tensile strain capacity ranging from 3.3% to 3.5%. It is suggested that the combination of SCMs (12.5% FA and 12.5% SL and the BA aggregate with the substitution rate of 10% can be effectively used in ECC preparation.

  1. Cementitious near-field sorption data bases for performance assessment of a L/ILW repository in a Palfris marl host rock. CEM-94: update I, June 1997

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M.H.; Loon, L.R. van [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1998-01-01

    This report is an update on an earlier cementitious sorption data base (SDB) prepared by Bradbury and Sarott (1994). The aim is to review any new information or data which have become available in the intervening time and modify the existing SDB appropriately. Discussions will be confined predominantly to areas which have led to significant changes to or reappraisals of the data/values or procedures for obtaining/modifying them. From this point of view this update and the previous SDB are closely related and belong together. The complexation of radionuclides with organic ligands from the chemical degradation of cellulose, and the subsequent negative effects on sorption properties, were identified as being processes of great importance. Since 1994 significant progress has been made in this field and a major part of this work is devoted to a reassessment of the impact of `organics` on near-field sorption. In particular, the very conservative assumptions which had been made previously because of the general lack of good quality data available at that time, could be replaced by realistic parameter estimates based on new knowledge. For example, maximum likely concentrations of cellulose degradation products and cement additives in the cement pore waters could be calculated allowing the potential effects of these organic ligands on sorption to be bounded. Sorption values for safety relevant radionuclides corresponding to the three broad stages of cement/concrete degradation during the lifetime of the repository are presented in tabulated form. The influence of the wide variety of organic ligands existing in the different waste categories, SMA-1 to SMA-4, is quantified in terms of sorption reduction factors. In the compilation of this cement SDB update, radionuclide uptake onto the vast quantities of aggregate materials and corrosion products from iron/steel was not taken into account. (author) 10 figs., 8 tabs., refs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2011-03-01

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

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

    Science.gov (United States)

    Samonov, A. E.

    2011-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-24

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1987-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Grutzeck, Michael W.

    2005-06-27

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  10. Porosimetry by DraMuTS applied to DEM-produced cementitious materials

    NARCIS (Netherlands)

    Stroeven, P.; Le, L.B.N.

    2013-01-01

    Porosimetry of matured cementitious materials is of hiigh interest, because durability is governed by the capillary pore network structure. Experimental approaches are time-consuming, laborious and thus expensive. Fast developments in computer technology make it nowadays possible quite realistically

  11. Porosimetry by DraMuTS applied to DEM-produced cementitious materials

    NARCIS (Netherlands)

    Stroeven, P.; Le, L.B.N.

    2013-01-01

    Porosimetry of matured cementitious materials is of hiigh interest, because durability is governed by the capillary pore network structure. Experimental approaches are time-consuming, laborious and thus expensive. Fast developments in computer technology make it nowadays possible quite realistically

  12. Predicting the Probability of Failure of Cementitious Sewer Pipes Using Stochastic Finite Element Method.

    Science.gov (United States)

    Alani, Amir M; Faramarzi, Asaad

    2015-06-10

    In this paper, a stochastic finite element method (SFEM) is employed to investigate the probability of failure of cementitious buried sewer pipes subjected to combined effect of corrosion and stresses. A non-linear time-dependant model is used to determine the extent of concrete corrosion. Using the SFEM, the effects of different random variables, including loads, pipe material, and corrosion on the remaining safe life of the cementitious sewer pipes are explored. A numerical example is presented to demonstrate the merit of the proposed SFEM in evaluating the effects of the contributing parameters upon the probability of failure of cementitious sewer pipes. The developed SFEM offers many advantages over traditional probabilistic techniques since it does not use any empirical equations in order to determine failure of pipes. The results of the SFEM can help the concerning industry (e.g., water companies) to better plan their resources by providing accurate prediction for the remaining safe life of cementitious sewer pipes.

  13. Long-term cement corrosion in chloride-rich solutions relevant to radioactive waste disposal in rock salt - Leaching experiments and thermodynamic simulations

    Science.gov (United States)

    Bube, C.; Metz, V.; Bohnert, E.; Garbev, K.; Schild, D.; Kienzler, B.

    Low- and intermediate-level radioactive wastes are frequently solidified in a cement matrix. In a potential repository for nuclear wastes, the cementitious matrix is altered upon contact with solution and the resulting secondary phases may provide for significant retention of the radionuclides incorporated in the wastes. In order to assess the secondary phases formed upon corrosion in chloride-rich solutions, which are relevant for nuclear waste disposal in rock salt, leaching experiments were performed. Conventional laboratory batch experiments using powdered hardened cement paste in MgCl2-rich solutions were left to equilibrate for up to three years and full-scale cemented waste products were exposed to NaCl-rich and MgCl2-rich solutions for more than twenty years, respectively. Solid phase analyses revealed that corrosion of hardened cement in MgCl2-rich solutions advanced faster than in NaCl-rich solutions due to the extensive exchange of Mg from solution against Ca from the cementitious solid. Thermodynamic equilibrium simulations compared well to results at the final stages of the respective experiments indicating that close to equilibrium conditions were reached. At high cement product to brine ratios (>0.65 g mL-1), the solution composition in the laboratory-scale experiments was close to that of the full-scale experiments (cement to brine ratio of 2.5 g mL-1) in the MgCl2 systems. The present study demonstrates the applicability of thermodynamic methods used in this approach to adequately describe full-scale long-term experiments with cemented waste simulates.

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

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

  15. Cementitious composite materials with improved self-healing potential

    Directory of Open Access Journals (Sweden)

    Cornelia BAERA

    2015-12-01

    Full Text Available Cement-based composites have proved, over the time, certain abilities of self-healing the damages (cracks and especially microcracs that occur within their structure. Depending on the level of damage and of the composite type in which this occurs, the self - healing process (SH can range from crack closing or crack sealing to the stage of partial or even complete recovery of material physical - mechanical properties. The aim of this paper is to present the general concept of Engineered Cementitious Composites (ECCs with their unique properties including their self-healing (SH capacity, as an innovative direction for a global sustainable infrastructure. The experimental steps initiated for the development in Romania of this unique category of materials, using materials available on the local market, are also presented.

  16. Talc-based cementitious products: Effect of talc calcination

    Directory of Open Access Journals (Sweden)

    C.J. Ngally Sabouang

    2015-09-01

    Full Text Available This study reports the use of calcined talc for cementitious products making. The calcination is used to enhance the availability of magnesium from talc to react with phosphate for cement phase formation. It is shown that previous calcination of talc leads to products having enhanced mechanical performance due to the formation of more cement phase than in products based on raw talc. Talc fired at 900 °C was found to be the one in which magnesium release was maximal. Firing at temperature higher than 900 °C leads to the stabilization of enstatite, which decreased the magnesium availability. The cement phase is struvite, which was better detected on the X-ray patterns of the products involving fired talc. All the products have very rapid setting time and low shrinkage.

  17. Dynamic fracture behaviour in fibre-reinforced cementitious composites

    Science.gov (United States)

    Yu, Rena C.; Cifuentes, Héctor; Rivero, Ignacio; Ruiz, Gonzalo; Zhang, Xiaoxin

    2016-08-01

    The object of this work is to simulate the dynamic fracture propagation in fibre-reinforced cementitious composites, in particular, in steel fibre reinforced concrete (SFRC). Beams loaded in a three-point bend configuration through a drop-weight impact device are considered. A single cohesive crack is assumed to propagate at the middle section; the opening of this crack is governed by a rate-dependent cohesive law; the fibres around the fracture plane are explicitly represented through truss elements. The fibre pull-out behaviour is depicted by an equivalent constitutive law, which is obtained from an analytical load-slip curve. The obtained load-displacement curves and crack propagation velocities are compared with their experimental counterparts. The good agreement with experimental data testifies to the feasibility of the proposed methodology and paves the way to its application in a multi-scale framework.

  18. STEREOLOGICAL ESTIMATES FOR ROUGHNESS AND TORTUOSITY IN CEMENTITIOUS COMPOSITES

    Directory of Open Access Journals (Sweden)

    Piet Stroeven

    2011-05-01

    Full Text Available Relatively weak interfaces between aggregate grains and the cementitious matrix initiate the damage evolution process leading to fracture. Coalescence between nearby interface cracks is promoted by the small nearest neighbour distances in a dense random packing of the aggregate. The fracture surface is therefore modelled as a dividing plane from which particles protrude. Assuming spherical aggregate, roughness is obtained as the global geometrical-statistical expression for the increase in fracture surface area due to a multitude of dome-like caps of various sizes. Transport phenomena in concrete are equally influenced by the aggregate, because traversing water-born molecules or ions have to go around the dense grains. This route is additionally promoted by the relatively high porosity in the interfacial transition zone. The planar and linear concepts of tortuosity in the transport path are analogous to those of roughness.

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

    Science.gov (United States)

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

    2015-04-15

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

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

    Science.gov (United States)

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

    2009-11-15

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