National Library of Energy BETA

Sample records for waste form selection

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

    SciTech Connect (OSTI)

    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.

  2. Secondary Waste Form Down Selection Data Package – Ceramicrete

    SciTech Connect (OSTI)

    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 binder is formed through an acid-base reaction between calcined magnesium oxide (MgO; a base) and potassium hydrogen phosphate (KH{sub 2}PO{sub 4}; an acid) in aqueous solution. The reaction product sets at room temperature to form a highly crystalline material. During the reaction, the hazardous and radioactive contaminants also react with KH{sub 2}PO{sub 4} to form highly insoluble phosphates. In this data package, physical property and waste acceptance data for Ceramicrete waste forms fabricated with wastes having compositions that were similar to those expected for secondary waste effluents, as well as secondary waste effluent simulants from the Hanford Tank Waste Treatment and Immobilization Plant were reviewed. With the exception of one secondary waste form formulation (25FA+25 W+1B.A. fabricated with the mixed simulant did not meet the compressive strength requirement), all the Ceramicrete waste forms that were reviewed met or exceeded Integrated Disposal Facility waste acceptance criteria.

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

    SciTech Connect (OSTI)

    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.

  4. A Strategy to Assess Performance of Selected Low-Activity Waste Forms in an Integrated Disposal Facility

    SciTech Connect (OSTI)

    McGrail, B PETER.; Bacon, Diana H.; Serne, R JEFFREY.; Pierce, Eric M.

    2003-08-22

    An overall strategy for evaluating the long-term performance of three waste forms being considered for supplemental treatment of low-activity waste at Hanford is discussed. The same computational framework used to conduct the 2001 ILAW performance assessment will be used for all three waste forms. Cast stone will be modeled with a diffusion-advection transport model and bulk vitrified glass and steam reformed LAW will be modeled with a reactive chemical transport simulator. The recommended laboratory testing to support the supplemental LAW form selection includes single-pass flow-through (SPFT), product consistency (PCT), and vapor hydration tests for glass, SPFT and PCT tests for steam reformed LAW forms, and ANS 16.1 tests for cast stone. These and potentially other laboratory tests for the selected waste form(s) would also be the basis for more detailed studies needed to support a comprehensive long-term performance assessment should one or more of these waste forms be selected for disposal in an integrated disposal facility.

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

    SciTech Connect (OSTI)

    Jantzen, C. M.; Pierce, E. M.; Bannochie, C. J.; Burket, P. R.; Cozzi, A. D.; Crawford, C. L.; Daniel, W. E.; Fox, K. M.; Herman, C. C.; Miller, D. H.; Missimer, D. M.; Nash, C. A.; Williams, M. F.; Brown, C. F.; Qafoku, N. P.; Neeway, J. J.; Valenta, M. M.; Gill, G. A.; Swanberg, D. J.; Robbins, R. A.; Thompson, L. E.

    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.

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

    SciTech Connect (OSTI)

    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 268.48. The data from the EPA draft 1315 leach test showed that LI values for COCs, namely 99Tc and I, ranged from 8.2 to 11.4 and 4.3 to 7.5, respectively. These values indicate that 99Tc meets the WAC LI requirement of 9.0 whereas, the LI values for I does not meet the WAC requirement of 11.0. Results of Toxicity Characteristic Leaching Procedure (TCLP)(EPA Method 1311) conducted on Waste Simulant S1-optimized DuraLith specimens, indicated that the concentrations of RCRA metals (Ag, As, Cd, Cr, Hg, and Pb) in the leachates were well below the Universal Treatment Standard limits in 40 CFR 268.48. The data from the ANSI/ANS 16.1 leach test showed that LI values for COC, namely Re (as a Tc surrogate), ranged from 8.06 to 10.81. The LI value for another COC, namely I, was not measured in this test. The results of the compressive strength testing of Waste Simulant S1-optimized DuraLith specimens indicated that the monoliths were physically robust with compressive strengths ranging from 115.5 MPa (16757 psi) to 156.2 MPA (22667 psi).

  7. Secondary Waste Cast Stone Waste Form Qualification Testing Plan

    SciTech Connect (OSTI)

    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

  8. Report on Intact and Degraded Criticality for Selected Plutonium Waste Forms in a Geologic Repository, Volume I: MOX SNF

    SciTech Connect (OSTI)

    J.A. McClure

    1998-09-21

    As part of the plutonium waste form development and down-select process, repository analyses have been conducted to evaluate the long-term performance of these forms for repository acceptance. Intact and degraded mode criticality analysis of the mixed oxide (MOX) spent fuel is presented in Volume I, while Volume II presents the evaluations of the waste form containing plutonium immobilized in a ceramic matrix. Although the ceramic immobilization development program is ongoing, and refinements are still being developed and evaluated, this analysis provides value through quick feed-back to this development process, and as preparation for the analysis that will be conducted starting in fiscal year (FY) 1999 in support of the License Application. While no MOX fuel has been generated in the United States using weapons-usable plutonium, Oak Ridge National Laboratory (ORNL) has conducted calculations on Westinghouse-type reactors to determine the expected characteristics of such a fuel. These spent nuclear fuel (SNF) characteristics have been used to determine the long-term potential for criticality in a repository environment. In all instances the methodology and scenarios used in these analyses are compatible with those developed and used for Commercial Spent Nuclear Fuel (CSNF) and Defense High Level Waste (DHLW), as tailored for the particular characteristics of the waste forms. This provides a common basis for comparison of the results. This analysis utilizes dissolution, solubility, and thermodynamic data that are currently available. Additional data on long-term behavior is being developed, and later analyses (FY 99) to support the License Application will use the very latest information that has been generated. Ranges of parameter values are considered to reflect sensitivity to uncertainty. Most of the analysis is focused on those parameter values that produce the worst case results, so that potential licensing issues can be identified.

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

    SciTech Connect (OSTI)

    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.

  10. CERAMIC WASTE FORM DATA PACKAGE

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  12. Miscellaneous Waste-Form FEPs

    SciTech Connect (OSTI)

    A. Schenker

    2000-12-08

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

  13. Waste Form Degradation Model Integration for Engineered Materials...

    Office of Environmental Management (EM)

    Waste Form Degradation Model Integration for Engineered Materials Performance Waste Form Degradation Model Integration for Engineered Materials Performance The collaborative...

  14. Secondary Waste Forms and Technetium Management

    Office of Environmental Management (EM)

    Secondary Waste Forms and Technetium Management Joseph H. Westsik, Jr. Pacific Northwest National Laboratory EM HLW Corporate Board Meeting November 18, 2010 What are Secondary...

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

    SciTech Connect (OSTI)

    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 all the waste forms had leachability indices better than the target LI > 9 for technetium; (2) Rhenium diffusivity: Cast Stone 2M specimens, when tested using EPA 1315 protocol, had leachability indices better than the target LI > 9 for technetium based on rhenium as a surrogate for technetium. All other waste forms tested by ANSI/ANS 16.1, ASTM C1308, and EPA 1315 test methods had leachability indices that were below the target LI > 9 for Tc based on rhenium release. These studies indicated that use of Re(VII) as a surrogate for 99Tc(VII) in low temperature secondary waste forms containing reductants will provide overestimated diffusivity values for 99Tc. Therefore, it is not appropriate to use Re as a surrogate 99Tc in future low temperature waste form studies. (3) Iodine diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that the three waste forms had leachability indices that were below the target LI > 11 for iodine. Therefore, it may be necessary to use a more effective sequestering material than silver zeolite used in two of the waste forms (Ceramicrete and DuraLith); (4) Sodium diffusivity: All the waste form specimens tested by the three leach methods (ANSI/ANS 16.1, ASTM C1308, and EPA 1315) exceeded the target LI value of 6; (5) All three leach methods (ANS 16.1, ASTM C1308 and EPA 1315) provided similar 99Tc diffusivity values for both short-time transient diffusivity effects as well as long-term ({approx}90 days) steady diffusivity from each of the three tested waste forms (Cast Stone 2M, Ceramicrete and DuraLith). Therefore, any one of the three methods can be used to determine the contaminant diffusivities from a selected waste form.

  16. DWPF waste form compliance plan (Draft Revision)

    SciTech Connect (OSTI)

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

    1991-01-01

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

  17. DWPF waste form compliance plan (Draft Revision)

    SciTech Connect (OSTI)

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

    1991-12-31

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

  18. DESCRIPTION OF SELECTED WASTE MANAGEMENT PROBLEMS,

    E-Print Network [OSTI]

    #12;DESCRIPTION OF SELECTED WASTE MANAGEMENT PROBLEMS, OPTIONS AND STRATEGIES Prepared for BC of Agriculture, Fisheries and Food Fisheries and Oceans Fraser River Action Plan November, 1996 Prepared by P. E Nutrients in Wastes 22 4.2.5 Waste Treatment 23 5.0 STRATEGY DEVELOPMENT 24 5.1 LAND USE MANAGEMENT 24 5

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

    SciTech Connect (OSTI)

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

    1982-09-01

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

  20. CRYSTALLINE CERAMIC WASTE FORMS: REFERENCE FORMULATION REPORT

    SciTech Connect (OSTI)

    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 with CS/LN/TM combined waste stream with Mo and Zr removed. Waste streams that contain Mo must be produced in reducing environments to avoid Cs-Mo oxide phase formation. Waste streams without Mo have the ability to be melt processed in air. A path forward for further optimizing the processing steps needed to form the targeted phase assemblages is outlined in this report. Processing modifications including melting in a reducing atmosphere, and controlled heat treatment schedules are anticipated to improve the targeted elemental partitioning.

  1. SEPARATIONS AND WASTE FORMS CAMPAIGN IMPLEMENTATION PLAN

    SciTech Connect (OSTI)

    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.

  2. Alternative Waste Forms for Electro-Chemical Salt Waste

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  4. DOE Selects Seven Contractors for Waste Treatment Basic Ordering...

    Office of Environmental Management (EM)

    Selects Seven Contractors for Waste Treatment Basic Ordering Agreements DOE Selects Seven Contractors for Waste Treatment Basic Ordering Agreements June 4, 2015 - 12:00pm Addthis...

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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 pathways for the potential reaction products. The phase equilibria and thermodynamics involving the intermediates in the decay process in this program will assist in selection of the best process for Cs or Sr immobilization. In addition, data from the study can be used to develop engineering solutions for potential process upsets. Second, the glass – crystal stability of multicomponent oxide phases that were representative silicates on this program is highly distinguishable for mother compounds and decay products, thus providing a fundamental understanding on the separate effects from chemistry and from radiation. Finally, we have developed a foundation for understanding chemistry-structure-energetics relationships in titanosilicates that can be used to develop more effective materials.

  7. Audit of Selected Hazardous Waste Remedial Actions Program Costs...

    Office of Environmental Management (EM)

    of Selected Hazardous Waste Remedial Actions Program Costs, ER-B-97-04 Audit of Selected Hazardous Waste Remedial Actions Program Costs, ER-B-97-04 Audit of Selected Hazardous...

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

    SciTech Connect (OSTI)

    Ebert, William; Pereira, Candido; Heltemes, Thad A.; Youker, Amanda; Makarashvili, Vakhtang; Vandegrift, George F.

    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.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

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

    1993-08-01

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

  11. Consolidation process for producing ceramic waste forms

    DOE Patents [OSTI]

    Hash, Harry C. (Joliet, IL); Hash, Mark C. (Shorewood, IL)

    2000-01-01

    A process for the consolidation and containment of solid or semisolid hazardous waste, which process comprises closing an end of a circular hollow cylinder, filling the cylinder with the hazardous waste, and then cold working the cylinder to reduce its diameter while simultaneously compacting the waste. The open end of the cylinder can be sealed prior to or after the cold working process. The preferred method of cold working is to draw the sealed cylinder containing the hazardous waste through a plurality of dies to simultaneously reduce the diameter of the tube while compacting the waste. This process provides a quick continuous process for consolidating hazardous waste, including radioactive waste.

  12. Laboratory procedures for waste form testing

    SciTech Connect (OSTI)

    Mast, E.S.

    1994-09-19

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

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

    SciTech Connect (OSTI)

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

    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 binder components from the waste form surface. Waste forms for ANS 16.1 leach testing contained appropriate amounts of rhenium and iodine as radionuclide surrogates, along with the additives silver-loaded zeolite and tin chloride. The leachability index for Re was found to range from 7.9 to 9.0 for all the samples evaluated. Iodine was below detection limit (5 ppb) for all the leachate samples. Further, leaching of sodium was low, as indicated by the leachability index ranging from 7.6-10.4, indicative of chemical binding of the various chemical species. Target leachability indices for Re, I, and Na were 9, 11, and 6, respectively. Degradation was observed in some of the samples post 90-day ANS 16.1 tests. Toxicity characteristic leaching procedure (TCLP) results showed that all the hazardous contaminants were contained in the waste, and the hazardous metal concentrations were below the Universal Treatment Standard limits. Preliminary scale-up (2-gal waste forms) was conducted to demonstrate the scalability of the Ceramicrete process. Use of minimal amounts of boric acid as a set retarder was used to control the working time for the slurry. Flexibility in treating waste streams with wide ranging compositional make-ups and ease of process scale-up are attractive attributes of Ceramicrete technology.

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

    SciTech Connect (OSTI)

    Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

    2014-05-09

    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 4,136 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.

  15. Challenges in Modeling the Degradation of Ceramic Waste Forms

    SciTech Connect (OSTI)

    Devanathan, Ramaswami; Gao, Fei; Sun, Xin

    2011-09-01

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

  16. Method for forming microspheres for encapsulation of nuclear waste

    DOE Patents [OSTI]

    Angelini, Peter (Oak Ridge, TN); Caputo, Anthony J. (Knoxville, TN); Hutchens, Richard E. (Knoxville, TN); Lackey, Walter J. (Oak Ridge, TN); Stinton, David P. (Knoxville, TN)

    1984-01-01

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

  17. Small businesses selected for nuclear waste services

    E-Print Network [OSTI]

    to the Waste Isolation Pilot Plant (WIPP) repository in Southern New Mexico. "These tasks play a key role of above-ground LANL transuranic waste to WIPP by June 30, 2014," said George Rael, Environmental Projects

  18. Process for immobilizing plutonium into vitreous ceramic waste forms

    DOE Patents [OSTI]

    Feng, X.; Einziger, R.E.

    1997-01-28

    Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

  19. Process for immobilizing plutonium into vitreous ceramic waste forms

    DOE Patents [OSTI]

    Feng, X.; Einziger, R.E.

    1997-08-12

    Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

  20. Process for immobilizing plutonium into vitreous ceramic waste forms

    DOE Patents [OSTI]

    Feng, Xiangdong (Richland, WA); Einziger, Robert E. (Richland, WA)

    1997-01-01

    Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

  1. Transuranic contaminated waste form characterization and data base

    SciTech Connect (OSTI)

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

    1980-07-01

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

  2. Waste Form Qualification Compliance Strategy for Bulk Vitrification

    SciTech Connect (OSTI)

    Bagaasen, Larry M.; Westsik, Joseph H.; Brouns, Thomas M.

    2005-01-03

    The Bulk Vitrification System is being pursued to assist in immobilizing the low-activity tank waste from the 53 million gallons of radioactive waste in the 177 underground storage tanks on the Hanford Site. To demonstrate the effectiveness of the bulk vitrification process, a research and development facility known as the Demonstration Bulk Vitrification System (DBVS) is being built to demonstrate the technology. Specific performance requirements for the final packaged bulk vitrification waste form have been identified. In addition to the specific product-performance requirements, performance targets/goals have been identified that are necessary to qualify the waste form but do not lend themselves to specifications that are easily verified through short-term testing. Collectively, these form the product requirements for the DBVS. This waste-form qualification (WFQ) strategy document outlines the general strategies for achieving and demonstrating compliance with the BVS product requirements. The specific objectives of the WFQ activities are discussed, the bulk vitrification process and product control strategy is outlined, and the test strategy to meet the WFQ objectives is described. The DBVS product performance targets/goals and strategies to address those targets/goals are described. The DBVS product-performance requirements are compared to the Waste Treatment and Immobilization Plant immobilized low-activity waste product specifications. The strategies for demonstrating compliance with the bulk vitrification product requirements are presented.

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

    SciTech Connect (OSTI)

    Holtzscheiter, E.W. [Westinghouse Savannah River Company, AIKEN, SC (United States); Harbour, J.R.

    1998-05-01

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

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

    SciTech Connect (OSTI)

    Stewart, Martin W.A.; Moricca, Sam A.; Zhang, Yingjie; Day, R. Arthur; Begg, Bruce D.; Scales, Charlie R.; Maddrell, Ewan R.; Hobbs, Jeff

    2013-07-01

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

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

    SciTech Connect (OSTI)

    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 conductivity, pH and analytical concentration measured as a function of time decrease exponentially. In some cases nitrate, sulfate, chloride and fluoride ion concentrations increased with time and processing temperature with respect to the reference sample. The increasing concentration of these ions was due to the lack of formation of crystalline phases that can incorporate them in their structures, especially cancrinite. Another plausible explanations for their increase was due to the continuous withdrawal of cations with time, for example sodium to form zeolites, thereby increase their concentrations.

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

    SciTech Connect (OSTI)

    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.).

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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 crack during cooling and crystals may be prone to dissolution. By designing a glass-ceramics, the risks of deleterious effects from devitrification are removed. Furthermore, glass-ceramics have higher mechanical strength and impact strengths and possess greater chemical durability as noted above. Glass-ceramics should provide a waste form with the advantages of glass - ease of manufacture - with improved mechanical properties, thermal stability, and chemical durability. This report will cover aspects relevant for the validation of the CCIM use in the production of glass-ceramic waste forms.

  9. Separations and Waste Forms Research and Development: FY 2012 Accomplishments Report

    SciTech Connect (OSTI)

    Not Listed

    2013-02-01

    This report contains FY 2012 accomplishments for the Separations and Waste Form Research and Development Project.

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

    SciTech Connect (OSTI)

    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.

  11. Waste form development for use with ORNL waste treatment facility sludge

    SciTech Connect (OSTI)

    Abotsi, G.M.K.; Bostick, W.D.

    1996-05-01

    A sludge that simulates Water Softening Sludge number 5 (WSS number 5 filtercake) at Oak Ridge National Laboratory was prepared and evaluated for its thermal behavior, volume reduction, stabilization, surface area and compressive strength properties. Compaction of the surrogate waste and the calcium oxide (produced by calcination) in the presence of paraffin resulted in cylindrical molds with various degrees of stability. This work has demonstrated that surrogate WSS number 5 at ORNL can be successfully stabilized by blending it with about 35 percent paraffin and compacting the mixture at 8000 psi. This compressive strength of the waste form is sufficient for temporary storage of the waste while long-term storage waste forms are developed. Considering the remarkable similarity between the surrogate and the actual filtercake, the findings of this project should be useful for treating the sludge generated by the waste treatment facility at ORNL.

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

    SciTech Connect (OSTI)

    Not Available

    1980-09-30

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

  13. Degradation modeling of the ANL ceramic waste form

    SciTech Connect (OSTI)

    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.

  14. The effect of concentration on the structure and crystallinity of a cementitious waste form for caustic wastes

    SciTech Connect (OSTI)

    Chung, Chul-Woo; Turo, Laura A.; Ryan, Joseph V.; Johnson, Bradley R.; McCloy, John S.

    2013-06-01

    Cement-based waste forms have long been considered economical technologies for disposal of various types of waste. A solidified cementitious waste form, Cast Stone, was developed to immobilize the radioactive secondary waste from vitrification processes. In this work, Cast Stone was considered for a Na-based caustic liquid waste, and its physical properties were analyzed as a function of liquid waste loading up to 2 M Na. Differences in crystallinity (phase composition), microstructure, mesostructure (pore size distribution, surface area), and macrostructure (density, compressive strength) were investigated using various analytical techniques, in order to assess the suitability of Cast Stone as a chemically durable waste. It was found that the concentration of secondary waste simulant (caustic waste) had little effect on the relevant engineering properties of Cast Stone, showing that Cast Stone could be an effective and tolerant waste form for a wide range of concentrations of high sodium waste.

  15. Surface effects of cement-based solidified waste forms 

    E-Print Network [OSTI]

    Pavlonnis, George

    1998-01-01

    This study was performed in order to determine-nine if the surface characteristics of cement-based waste forms were different than those of the bulk material. This was done as a prelude to the potential development of an accelerated leach test...

  16. Method of making nanostructured glass-ceramic waste forms

    SciTech Connect (OSTI)

    Gao, Huizhen; Wang, Yifeng; Rodriguez, Mark A.; Bencoe, Denise N.

    2014-07-08

    A waste form for and a method of rendering hazardous materials less dangerous is disclosed that includes fixing the hazardous material in nanopores of a nanoporous material, reacting the trapped hazardous material to render it less volatile/soluble, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.

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

    SciTech Connect (OSTI)

    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.

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

    DOE Patents [OSTI]

    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.

  19. Transuranic contaminated waste form characterization and data base

    SciTech Connect (OSTI)

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

    1980-07-01

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

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

    SciTech Connect (OSTI)

    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.

  1. Low-level radioactive waste technology: a selected, annotated bibliography

    SciTech Connect (OSTI)

    Fore, C.S.; Vaughan, N.D.; Hyder, L.K.

    1980-10-01

    This annotated bibliography of 447 references contains scientific, technical, economic, and regulatory information relevant to low-level radioactive waste technology. The bibliography focuses on environmental transport, disposal site, and waste treatment studies. The publication covers both domestic and foreign literature for the period 1952 to 1979. Major chapters selected are Chemical and Physical Aspects; Container Design and Performance; Disposal Site; Environmental Transport; General Studies and Reviews; Geology, Hydrology and Site Resources; Regulatory and Economic Aspects; Transportation Technology; Waste Production; and Waste Treatment. Specialized data fields have been incorporated into the data file to improve the ease and accuracy of locating pertinent references. Specific radionuclides for which data are presented are listed in the Measured Radionuclides field, and specific parameters which affect the migration of these radionuclides are presented in the Measured Parameters field. In addition, each document referenced in this bibliography has been assigned a relevance number to facilitate sorting the documents according to their pertinence to low-level radioactive waste technology. The documents are rated 1, 2, 3, or 4, with 1 indicating direct applicability to low-level radioactive waste technology and 4 indicating that a considerable amount of interpretation is required for the information presented to be applied. The references within each chapter are arranged alphabetically by leading author, corporate affiliation, or title of the document. Indexes are provide for (1) author(s), (2) keywords, (3) subject category, (4) title, (5) geographic location, (6) measured parameters, (7) measured radionuclides, and (8) publication description.

  2. INITIAL SELECTION OF SUPPLEMENTAL TREATMENT TECHNOLOGIES FOR HANFORDS LOW ACTIVITY TANK WASTE

    SciTech Connect (OSTI)

    RAYMOND, R.E.

    2004-02-20

    In 2002, the U.S. Department of Energy (DOE) documented a plan for accelerating cleanup of the Hanford Site, located in southeastern Washington State, by at least 35 years. A key element of the plan was acceleration of the tank waste program and completion of ''tank waste treatment by 2028 by increasing the capacity of the planned Waste Treatment Plant (WTP) and using supplemental technologies for waste treatment and immobilization.'' The plan identified specific technologies to be evaluated for supplemental treatment of as much as 70% of the low-activity waste (LAW). In concert with this acceleration plan, DOE, the U.S. Environmental Protection Agency, and the Washington State Department of Ecology proposed to accelerate--from 2014 to 2006--the Hanford Federal Facility Agreement and Consent Order milestone (M-62-11) associated with a final decision on the balance of tank waste that is beyond the capacity of the WTP. The DOE Office of River Protection tank farm contractor, CH2M HILL Hanford Group, Inc. (CH2M HILL), was tasked with testing and evaluating selected supplemental technologies to support final decisions on tank waste treatment. Three technologies and corresponding vendors were selected to support an initial technology selection in 2003. The three technologies were containerized grout called cast stone (Fluor Federal Services); bulk vitrification (AMEC Earth and Environmental, Inc.); and steam reforming (THOR Treatment Technologies, LLC.). The cast stone process applies an effective grout waste formulation to the LAW and places the cement-based product in a large container for solidification and disposal. Unlike the WTP LAW treatment, which applies vitrification within continuous-fed joule-heated ceramic melters, bulk vitrification produces a glass waste form using batch melting within the disposal container. Steam reforming produces a granular denitrified mineral waste form using a high-temperature fluidized bed process. An initial supplemental technology selection was completed in December 2003, enabling DOE and CH2M HILL to focus investments in 2004 on the testing and production-scale demonstrations needed to support the 2006 milestone.

  3. Reevaluation Of Vitrified High-Level Waste Form Criteria For Potential Cost Savings At The Defense Waste Processing Facility

    SciTech Connect (OSTI)

    Ray, J. W.; Marra, S. L.; Herman, C. C.

    2013-01-09

    At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form.

  4. Reevaluation of Vitrified High-Level Waste Form Criteria for Potential Cost Savings at the Defense Waste Processing Facility - 13598

    SciTech Connect (OSTI)

    Ray, J.W. [Savannah River Remediation (United States)] [Savannah River Remediation (United States); Marra, S.L.; Herman, C.C. [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)] [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01

    At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form. (authors)

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

    SciTech Connect (OSTI)

    McDonell, W R

    1982-04-01

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

  6. Cold Crucible Induction Melter Studies for Making Glass Ceramic Waste Forms: A Feasibility Assessment

    SciTech Connect (OSTI)

    Crum, Jarrod V.; Maio, Vincent; McCloy, John S.; Scott, Clark; Riley, Brian J.; Benefiel, Bradley; Vienna, John D.; Archibald, Kip; Rodriguez, Carmen P.; Rutledge, Veronica; Zhu, Zihua; Ryan, Joseph V.; Olszta, Matthew J.

    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 meter (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.

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

    SciTech Connect (OSTI)

    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 (approximately 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.

  8. Dissolution testing of a metallic waste form in chloride brine

    SciTech Connect (OSTI)

    Dawn E Janney

    2006-11-01

    This paper is intended for publication in the peer-reviewed proceedings from the Scientific Basis for Nuclear Waste Management (at the Fall 2006 meeting of the Materials Research Society). The same material was presented in a 15-minute talk. Argonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel from the Experimental Breeder Reactor II (EBR-II). One waste stream from this process consists of a metal waste form (MWF) whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr) and whose microstructure is a eutectic intergrowth of iron solid solutions and Fe-Zr-Cr-Ni intermetallics. This paper reports scanning electron microscope (SEM) observations of corrosion products formed during static immersion tests in which coupons of surrogate MWF containing 10 wt% U (SS-15Zr-10U) were immersed in solutions with nominal pH values of 3 and 4 and 1000 ppm added chloride for 70 days at 50 °C. Although the majority of the surface areas of the coupons appear unchanged, linear areas with localized corrosion products apparently consisting of porous materials overlying corrosion-product-filled channels formed on both coupons, cross-cutting phase boundaries in the original eutectic microstructures. Many of the linear areas intersected the sample edge at notches present before the tests or followed linear flaws visible in pre-test images. Compositions of corrosion products differed significantly from the bulk composition, and the maximum observed concentration of U in corrosion products (~25 at%) slightly exceeded the highest reported values in actinide-bearing phases in uncorroded surrogate MWF samples with comparable concentrations of U (~17-19 at%).

  9. Crystalline Ceramic Waste Forms: Comparison Of Reference Process For

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfate Reducing BacteriaConnect Collider TestspolycarbonateArticle)allostericCeramic Waste Form

  10. Radioactive waste management treatments: A selection for the Italian scenario

    SciTech Connect (OSTI)

    Locatelli, G. [Univ. of Lincoln, Lincoln School of Engineering, Brayford Pool - Lincoln LN6 7TS (United Kingdom); Mancini, M. [Politecnico di Milano, Dept. of Management, Economics and Industrial Engineering, Via Lambruschini 4/B, Milano (Italy); Sardini, M. [Politecnico di Milano, Dept. of Energy, Via Lambruschini 4, Milano (Italy)

    2012-07-01

    The increased attention for radioactive waste management is one of the most peculiar aspects of the nuclear sector considering both reactors and not power sources. The aim of this paper is to present the state-of-art of treatments for radioactive waste management all over the world in order to derive guidelines for the radioactive waste management in the Italian scenario. Starting with an overview on the international situation, it analyses the different sources, amounts, treatments, social and economic impacts looking at countries with different industrial backgrounds, energetic policies, geography and population. It lists all these treatments and selects the most reasonable according to technical, economic and social criteria. In particular, a double scenario is discussed (to be considered in case of few quantities of nuclear waste): the use of regional, centralized, off site processing facilities, which accept waste from many nuclear plants, and the use of mobile systems, which can be transported among multiple nuclear sites for processing campaigns. At the end the treatments suitable for the Italian scenario are presented providing simplified work-flows and guidelines. (authors)

  11. Progress in the Assessment of Waste-forms for the Immobilisation of UK Civil Plutonium

    SciTech Connect (OSTI)

    Harrison, M.T.; Scales, C.R.; Maddrell, E.R. [Nexia Solutions Ltd., British Technology Centre, Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom)

    2008-07-01

    The alternatives for the disposition of the UK's civil plutonium stocks are currently being investigated by Nexia Solutions Ltd. on behalf of the Nuclear Decommissioning Authority (NDA). A number of scenarios are currently being considered depending on the strategic requirements of the UK. The two main disposition options are: re-use as MOX (Mixed Oxide) fuel in reactors, or immobilisation in the event of any material being declared surplus to requirements. The amount of Pu which will require immobilisation will depend on future UK nuclear strategy, along with the extent of any stocks deemed unsuitable for re-use. However, it is likely that some portion will have to be immobilised and therefore three credible waste-forms are under consideration; ceramic, glass and 'immobilisation' MOX. These are currently being developed and assessed in a systematic programme that involves periodic evaluation against a range of criteria. In this way, by down-selecting on the basis of robust and technical review, the most appropriate option for immobilising surplus civil plutonium in the UK can be recommended. The latest results from the immobilisation experimental programme are presented following the de-selection of the least favourable glass and ceramic candidates. The main criteria for this decision were waste loading, durability, processability, criticality and proliferation resistance. In addition, the durability of unirradiated MOX fuel is being examined to determine its potential as a wasteform for Pu, and recent leach test data is discussed. The current evaluation comprises not only a comparison of the relevant physical properties of the various waste-forms, but also key processing parameters, e.g. glass viscosity and melter technology, ceramic fabrication routes, and criticality issues. Other important aspects of the long-term behaviour of the waste-forms under consideration in a potential repository environment, such as radiation damage, criticality control and the properties of any neutron poisons present, are also included. (authors)

  12. DIRECT DISPOSAL OF A RADIOACTIVE ORGANIC WASTE IN A CEMENTITIOUS WASTE FORM

    SciTech Connect (OSTI)

    Zamecnik, J; Alex Cozzi, A; Russell Eibling, R; Jonathan Duffey, J; Kim Crapse, K

    2007-02-22

    The disposition of {sup 137}Cs-containing tetraphenylborate (TPB) waste at the Savannah River Site (SRS) by immobilization in the cementitious waste form, or grout called ''saltstone'' was proposed as a straightforward, cost-effective method for disposal. Tests were performed to determine benzene release due to TPB decomposition in saltstone at several initial TPB concentrations and temperatures. The benzene release rates for simulants and radioactive samples were generally comparable at the same conditions. Saltstone monoliths with only the top surface exposed to air at 25 and 55 C at any tetraphenylborate concentration or at any temperature with 30 mg/L TPB gave insignificant releases of benzene. At higher TPB concentrations and 75 and 95 C, the benzene release could result in exceeding the Lower Flammable Limit in the saltstone vaults.

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

    SciTech Connect (OSTI)

    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 date and how they compare to testing performed on LAW glasses. Other details about vitreous waste form durability and impacts of REDuction/OXidation (REDOX) on durability are given in Appendix A. Details about the FBSR process, various pilot scale demonstrations, and applications are given in Appendix B. Details describing all the different leach tests that need to be used jointly to determine the leaching mechanisms of a waste form are given in Appendix C. Cautions regarding the way in which the waste form surface area is measured and in the choice of leachant buffers (if used) are given in Appendix D.

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

    SciTech Connect (OSTI)

    Lori Braase

    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.

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

    SciTech Connect (OSTI)

    Jiang, Weilin; Van Ginhoven, Renee M.

    2012-09-28

    This is a technical report summarizing the experimental and theoretical results for model waste form of aluminosilicate pollucite, obtained from January to September, 2012.

  16. 10/2/2006 SLAC-I-760-2A08Z-001-R002 HAZARDOUS WASTE DETERMINATION FORM

    E-Print Network [OSTI]

    Wechsler, Risa H.

    /2/2006 SLAC-I-760-2A08Z-001-R002 HAZARDOUS WASTE DETERMINATION FORM For RP Use Only Hazardous Waste;________________________________________________________________________________________________ 10/2/2006 SLAC-I-760-2A08Z-001-R002 HAZARDOUS WASTE DETERMINATION FORM For RP Use Only Hazardous Waste Codes:Hazardous Classification: [ ] Non-Hazardous [ ] RCRA Waste [ ] Non-RCRA Waste (CA Haz Waste

  17. Secondary Waste Form Development and Optimization—Cast Stone

    SciTech Connect (OSTI)

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

    SciTech Connect (OSTI)

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

    1982-06-01

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

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

    SciTech Connect (OSTI)

    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.

  20. Transuranic contaminated waste form characterization and data base

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Not Available

    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.

  2. NEAMS Nuclear Waste Management IPSC : evaluation and selection of tools for the quality environment.

    SciTech Connect (OSTI)

    Bouchard, Julie F.; Stubblefield, William Anthony; Vigil, Dena M.; Edwards, Harold Carter (Org. 1444 : Multiphysics Simulation Technology)

    2011-05-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Nuclear Waste Management Integrated Performance and Safety Codes (NEAMS Nuclear Waste Management IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. These M&S capabilities are to be managed, verified, and validated within the NEAMS Nuclear Waste Management IPSC quality environment. M&S capabilities and the supporting analysis workflow and simulation data management tools will be distributed to end-users from this same quality environment. The same analysis workflow and simulation data management tools that are to be distributed to end-users will be used for verification and validation (V&V) activities within the quality environment. This strategic decision reduces the number of tools to be supported, and increases the quality of tools distributed to end users due to rigorous use by V&V activities. This report documents an evaluation of the needs, options, and tools selected for the NEAMS Nuclear Waste Management IPSC quality environment. The objective of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation Nuclear Waste Management Integrated Performance and Safety Codes (NEAMS Nuclear Waste Management IPSC) program element is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to assess quantitatively the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. This objective will be fulfilled by acquiring and developing M&S capabilities, and establishing a defensible level of confidence in these M&S capabilities. The foundation for assessing the level of confidence is based upon the rigor and results from verification, validation, and uncertainty quantification (V&V and UQ) activities. M&S capabilities are to be managed, verified, and validated within the NEAMS Nuclear Waste Management IPSC quality environment. M&S capabilities and the supporting analysis workflow and simulation data management tools will be distributed to end-users from this same quality environment. The same analysis workflow and simulation data management tools that are to be distributed to end-users will be used for verification and validation (V&V) activities within the quality environment. This strategic decision reduces the number of tools to be supported, and increases the quality of tools distributed to end users due to rigorous use by V&V activities. NEAMS Nuclear Waste Management IPSC V&V and UQ practices and evidence management goals are documented in the V&V Plan. This V&V plan includes a description of the quality environment into which M&S capabilities are imported and V&V and UQ activities are managed. The first phase of implementing the V&V plan is to deploy an initial quality environment through the acquisition and integration of a set of software tools. An evaluation of the needs, options, and tools selected for the quality environment is given in this report.

  3. Selected radionuclides important to low-level radioactive waste management

    SciTech Connect (OSTI)

    1996-11-01

    The purpose of this document is to provide information to state representatives and developers of low level radioactive waste (LLW) management facilities about the radiological, chemical, and physical characteristics of selected radionuclides and their behavior in the environment. Extensive surveys of available literature provided information for this report. Certain radionuclides may contribute significantly to the dose estimated during a radiological performance assessment analysis of an LLW disposal facility. Among these are the radionuclides listed in Title 10 of the Code of Federal Regulations Part 61.55, Tables 1 and 2 (including alpha emitting transuranics with half-lives greater than 5 years). This report discusses these radionuclides and other radionuclides that may be significant during a radiological performance assessment analysis of an LLW disposal facility. This report not only includes essential information on each radionuclide, but also incorporates waste and disposal information on the radionuclide, and behavior of the radionuclide in the environment and in the human body. Radionuclides addressed in this document include technetium-99, carbon-14, iodine-129, tritium, cesium-137, strontium-90, nickel-59, plutonium-241, nickel-63, niobium-94, cobalt-60, curium -42, americium-241, uranium-238, and neptunium-237.

  4. Development of long-term performance models for radioactive waste forms

    SciTech Connect (OSTI)

    Bacon, Diana H.; Pierce, Eric M.

    2011-03-22

    The long-term performance of solid radioactive waste is measured by the release rate of radionuclides into the environment, which depends on corrosion or weathering rates of the solid waste form. The reactions involved depend on the characteristics of the solid matrix containing the radioactive waste, the radionuclides of interest, and their interaction with surrounding geologic materials. This chapter describes thermo-hydro-mechanical and reactive transport models related to the long-term performance of solid radioactive waste forms, including metal, ceramic, glass, steam reformer and cement. Future trends involving Monte-Carlo simulations and coupled/multi-scale process modeling are also discussed.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

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

    1983-02-25

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

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Wall, Nathalie A.; Neeway, James J.; Qafoku, Nikolla P.; Ryan, Joseph V.

    2015-09-30

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

  10. Report of Waste Discharge application (Form 200) | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onRAPID/Geothermal/Exploration/ColoradoRemsenburg-Speonk, New York: Energy ResourcesProductsRensselaerNorth Dakota:Waste

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

    SciTech Connect (OSTI)

    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. FLUIDIZED BED STEAM REFORMED MINERAL WASTE FORMS: CHARACTERIZATION AND DURABILITY TESTING

    SciTech Connect (OSTI)

    Jantzen, C; Troy Lorier, T; John Pareizs, J; James Marra, J

    2006-12-06

    Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium low activity wastes (LAW) such as those existing at the Hanford site, at the Idaho National Laboratory (INL), and the Savannah River Site (SRS). The addition of clay, charcoal, and a catalyst as co-reactants with the waste denitrates the aqueous wastes and forms a granular mineral waste form that can subsequently be made into a monolith for disposal if necessary. The waste form produced is a multiphase mineral assemblage of Na-Al-Si (NAS) feldspathoid minerals with cage and ring structures and iron bearing spinel minerals. The mineralization occurs at moderate temperatures between 650-750 C in the presence of superheated steam. The cage and ring structured feldspathoid minerals atomically bond radionuclides like Tc-99 and Cs-137 and anions such as SO{sub 4}, I, F, and Cl. The spinel minerals stabilize Resource Conservation and Recovery Act (RCRA) hazardous species such as Cr and Ni. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium bearing waste (SBW) in pilot scale facilities at the Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The results of the SPFT testing and the activation energies for dissolution are discussed in this study.

  13. FLUIDIZED BED STEAM REFORMED MINERAL WASTE FORMS: CHARACTERIZATION AND DURABILITY TESTING

    SciTech Connect (OSTI)

    Jantzen, C; Troy Lorier, T; John Pareizs, J; James Marra, J

    2007-03-31

    Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium low activity wastes (LAW) such as those existing at the Hanford site, at the Idaho National Laboratory (INL), and the Savannah River Site (SRS). The addition of clay, charcoal, and a catalyst as co-reactants with the waste denitrates the aqueous wastes and forms a granular mineral waste form that can subsequently be made into a monolith for disposal if necessary. The waste form produced is a multiphase mineral assemblage of Na-Al-Si (NAS) feldspathoid minerals with cage and ring structures and iron bearing spinel minerals. The mineralization occurs at moderate temperatures between 650-750 C in the presence of superheated steam. The cage and ring structured feldspathoid minerals atomically bond radionuclides like Tc-99 and Cs-137 and anions such as SO4, I, F, and Cl. The spinel minerals stabilize Resource Conservation and Recovery Act (RCRA) hazardous species such as Cr and Ni. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium bearing waste (SBW) in pilot scale facilities at the Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The results of the SPFT testing and the activation energies for dissolution are discussed in this study.

  14. Waste Isolation Pilot Plant Electronic FOIA Request Form

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home RoomPreservationBio-Inspired SolarAbout /Two0Photos and Videos/01/2012 Page 1 of 10 Waste

  15. Crystalline Ceramic Waste Forms: Comparison Of Reference Process For

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfate Reducing BacteriaConnect Collider TestspolycarbonateArticle)allostericCeramic Waste

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

    SciTech Connect (OSTI)

    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.

  17. Epsilon Metal Waste Form for Immobilization of Noble Metals from Used Nuclear Fuel

    SciTech Connect (OSTI)

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

    2013-10-01

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

  18. Special waste-form lysimeters - arid: 1984--1992 data summary and preliminary interpretation

    SciTech Connect (OSTI)

    Jones, T.L.; Serne, R.J.

    1994-10-01

    A lysimeter facility constructed at the Hanford Site in south-central Washington State has been used since 1984 to monitor the leaching of buried waste forms under natural conditions. The facility is generating data that are useful in evaluating source-term models used in radioactive waste transport analyses. The facility includes ten bare-soil lysimeters (183 cm diameter by 305 cm depth) containing buried waste forms generated at nuclear reactors in the United States and solidified with Portland M cement, masonry cement, bitumen, and vinyl-ester styrene. The waste forms contained in the lysimeters have been leached under natural, semiarid conditions. In spite of the semiarid conditions, from 1984 through 1992, an average of 45 cm of water leached through the lysimeters, representing 27% of area precipitation. Leachate samples have been routinely collected and analyzed for radionuclide and chemical content. To date, tritium, cobalt-60, and cesium-137 have been identified in the lysimeter leachate samples. From 1984 through 1992, over 4000 {mu}Ci of tritium, representing 76 and 71 % of inventory (not decay corrected), have been leached from the two waste forms containing tritium. Cobalt-60 has been found in the leachate from all six of the waste forms that originally contained > 1 mCi of inventory. The leached amounts of cobalt-60 represent < 0.1 % of original cobalt inventories. Mobile cobalt is believed to be chelated with organic compounds, such as ethylenediaminetetraacetic acid (EDTA), that are present in the waste. Trace amounts of cesium-137 have occasionally been identified in leachate from two waste forms since 1991. Qualitatively, the field leaching results confirm laboratory studies suggesting that tritium is readily leached from cement, and that cobalt-60 is generally leached more easily from cement than from vinyl-ester styrene.

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

    SciTech Connect (OSTI)

    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 demonstration project will provide additional options for fission product immobilization and waste management associated the electrochemical/pyrometallurgical processing of used nuclear fuel.

  20. 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

    SciTech Connect (OSTI)

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

    2012-02-02

    The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as {sup 137}Cs, {sup 129}I, {sup 99}Tc, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750 C) continuous method by which WTP-SW wastes can be processed irrespective of whether they 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 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.

  1. Ion Selective Ceramics for Waste Separations. Input for Annual Accomplishments Report

    SciTech Connect (OSTI)

    Spoerke, Erik David

    2015-10-01

    This report discusses“Ion-Selective Ceramics for Waste Separations” which aims to develop an electrochemical approach to remove fission product waste (e.g., Cs+ ) from the LiCl-KCl molten salts used in the pyroprocessing of spent nuclear fuel.

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

    SciTech Connect (OSTI)

    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 candidates, those of glass-ceramics (devitrified matrices) represent the best compromise for meeting the probable stricter disposal requirements in the future.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Bacon, Diana H.; McGrail, B PETER.

    2005-07-26

    A set of reactive chemical transport calculations was conducted with the Subsurface Transport Over Reactive Multiphases (STORM) code to evaluate the long-term performance of a representative low-activity waste glass in a shallow subsurface disposal system located on the Hanford Site. Two-dimensional simulations were run until the waste form release rates reached a quasi-stationary-state, usually after 2,000 to 4,000 yr. The primary difference between the waste form release simulations for the 2001 ILAW PA, and the simulations described herein, is the number of different materials considered. Whereas the previous PA considered only LAWABP1 glass, the current PA also describes radionuclide release from three different WTP glasses (LAWA44, LAWB45 and LAWC22), two different bulk vitrification glasses (6-tank composite and S-109), and three different grout waste forms (containing Silver Iodide, Barium Iodide and Barium Iodate). All WTP and bulk vitrification glasses perform well. However, the radionuclide release from the salt in the cast refractory surrounding the bulk vitrification waste packages is 2 to 170 times higher than the glass release rate, depending on the water recharge rate. Iodine-129 release from grouted waste forms is highly sensitive to the solubility of the iodine compound contained in the grout. The normalized iodine release rate from grout containing barium iodate is a factor of 10 higher than what the normalized release rate would be if the iodine were contained in LAWA44 glass.

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

    SciTech Connect (OSTI)

    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.

  6. Selection of Pretreatment Processes for Removal of Radionuclides from Hanford Tank Waste

    SciTech Connect (OSTI)

    CARREON, R.

    2002-01-01

    The U.S. Department of Energy's (DOE's), Office of River Protection (ORP) located at Hanford Washington has established a contract (1) to design, construct, and commission a new Waste Treatment and Immobilization Plant (WTP) that will treat and immobilize the Hanford tank wastes for ultimate disposal. The WTP is comprised of four major elements, pretreatment, LAW immobilization, HLW immobilization, and balance of plant facilities. This paper describes the technologies selected for pretreatment of the LAW and HLW tank wastes, how these technologies were selected, and identifies the major technology testing activities being conducted to finalize the design of the WTP.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  10. Paint selection for coating radioactive-waste drums

    SciTech Connect (OSTI)

    Briggs, J.L.

    1980-07-01

    It is concluded that although the white epoxy Paint Sample E is suitable for coating waste drums, the additional pretreated costs of grit blasting prior to paint application would preclude adoption of that paint system. The specified 10.0-mil coating thickness of that coating would also incur higher costs. The Vorac epoxy-phenolic base paint (buff or yellow) was the only other paint that exhibited suitable corrosion and impact resistance required for coating the waste drums. In addition, that paint does not require a grit-blasted substrate or other costly pretreatment prior to coating.

  11. Assessment of the Cast Stone Low-Temperature Waste Form Technology Coupled with Technetium Removal - 14379

    SciTech Connect (OSTI)

    Brown, Christopher F.; Rapko, Brian M.; Serne, R. Jeffrey; Westsik, Joseph H.; Cozzi, Alex; Fox, Kevin M.; Mccabe, Daniel J.; Nash, C. A.; Wilmarth, William R.

    2014-03-03

    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) were 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 and technetium removal projects at Hanford. Science and technology gaps 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 separation of technetium from waste processing streams. Technical approaches to address the science and technology gaps were identified and an initial sequencing priority was suggested. A subset of research was initiated in 2013 to begin addressing the most significant science and technology gaps. The purpose of this paper is to report progress made towards closing these gaps and provide notable highlights of results achieved to date.

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

    SciTech Connect (OSTI)

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

    2013-04-01

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

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

    SciTech Connect (OSTI)

    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 volume has expanded slightly by 0.3% again, presumably due to alpha-decay damage. (5) No bulk sample swelling was observed. (6) No amorphization of sodalite or actinide bearing phases was observed after four years of alpha-decay damage. (7) No microcracks or phase de-bonding were observed in waste form samples aged for four years. (8) In some areas of the {sup 238}Pu doped ceramic waste form material bubbles and voids were found. Bubbles and voids with similar size and density were also found in ceramic waste form samples without actinide. These bubbles and voids are interpreted as pre-existing defects. However, some contribution to these bubbles and voids from helium gas can not be ruled out. (9) Chemical durability of {sup 238}Pu CWF has not changed significantly after four years of alpha-decay exposure except for an increase in the release of salt components and Pu. Still, the plutonium release from CWF is very low at less than 0.005 g/m{sup 2}.

  14. DOE Selects Savannah River Remediation, LLC for Liquid Waste...

    Energy Savers [EERE]

    of Environmental Management and the Savannah River Site. Media contact(s): Kristin Brown, (202) 586-4940 Jim Giusti, (803) 952-7684 Addthis Related Articles DOE Selects...

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

    SciTech Connect (OSTI)

    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.

  16. Selectivity and development of the visual word form area

    E-Print Network [OSTI]

    King, Li-Wei

    2013-01-01

    An area of left occipitotemporal cortex commonly referred to as the visual word form area (VWFA), has consistently been shown to activate during the processing of written language. However, the exact nature of the region's ...

  17. Characterization of host phases for actinides in simulated metallic waste forms by transmission electron microscopy.

    SciTech Connect (OSTI)

    Janney, D. E.

    2005-11-21

    Argonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel prior to disposal. A waste stream from this process consists of stainless steel cladding hulls that contain undissolved metal fission products such as Tc, Ru, Rh, Pd, and Ag; a small amount of undissolved actinides (U, Np, Pu) also remains with the hulls. These wastes will be immobilized in a waste form whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr). Scanning electron microscope (SEM) observations of simulated metal waste forms (SS-15Zr with added actinides) show eutectic intergrowths of iron solid-solution (''steel'') and Fe-Zr-Cr-Ni (''intermetallic'') materials. The actinide elements are almost entirely in the intermetallic materials, where they occur in concentrations as high as 20 at%. Neutron- and electron-diffraction studies of the simulated waste forms show materials with structures similar to those of Fe{sub 2}Zr and Fe{sub 23}Zr{sub 6}. New TEM observations of simulated waste form samples with compositions SS-15Zr-2Np, SS-15Zr-5U, SS-15Zr-11U-0.6Ru-0.3Tc-0.1Pd, and SS-15Zr-10Pu suggest that the major U- and Pu-bearing phase has a structure similar to that of the C15 (cubic, MgCu{sub 2}-type) polymorph of Fe{sub 2}Zr. Materials with this structure exhibit significant variability in chemical compositions and actinide concentrations up to 20 at% (normalized so that atomic fractions of Cr, Ni, Fe, and Zr add up to 1). A U-bearing material similar to the C36 (dihexagonal, MgNi{sub 2}-type) polymorph of Fe{sub 2}Zr was also observed. Chemical variability in materials with the C36 Fe{sub 2}Zr structure is smaller than in those with the C15 Fe{sub 2}Zr structure, and U concentrations are less than 5 at%. Uranium concentrations up to 5 at.% were observed in materials with the Fe{sub 23}Zr{sub 6} (cubic, Mn{sub 23}Th{sub 6}-type) structure. Microstructures similar to those produced during experimental deformation of Fe-10 at% Zr alloys were observed in intermetallic materials in all of the simulated waste form samples. Stacking faults and associated dislocations are common in samples with U, but rarely observed in those with Np and Pu, while twins occur in all samples. Previously reported differences in dissolution behavior between samples with different actinides may be related to increased defect-assisted dissolution in samples with U.

  18. Evaluation of zirconium-iron-rhenium alloys as surrogates for a technetium alloy waste form 

    E-Print Network [OSTI]

    Mews, Paul Aaron

    2009-05-15

    and a second repository has not been selected. However, reprocessing technologies are currently under development to separate and reuse the energy-rich actinide elements from spent nuclear fuel. After reactor operation the vast majority of the spent... radioactive and some of then are strong neutron absorbers. These isotopes must be removed from the actinides and treated as high level nuclear waste. Reprocessing technology uses chemical separation methods to segregate the various constituents of spent...

  19. Multi-phase glass-ceramics as a waste form for combined fission products: alkalis, alkaline earths, lanthanides, and transition metals

    SciTech Connect (OSTI)

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

    2012-04-01

    In this study, multi-phase silicate-based glass-ceramics were investigated as an alternate waste form for immobilizing non-fissionable products from used nuclear fuel. Currently, borosilicate glass is the waste form selected for immobilization of this waste stream, however, the low thermal stability and solubility of MoO{sub 3} in borosilicate glass translates into a maximum waste loading in the range of 15-20 mass%. Glass-ceramics provide the opportunity to target durable crystalline phases, e.g., powellite, oxyapatite, celsian, and pollucite, that will incorporate MoO{sub 3} as well as other waste components such as lanthanides, alkalis, and alkaline earths at levels 2X the solubility limits of a single-phase glass. In addition a glass-ceramic could provide higher thermal stability, depending upon the properties of the crystalline and amorphous phases. Glass-ceramics were successfully synthesized at waste loadings of 42, 45, and 50 mass% with the following glass additives: B{sub 2}O{sub 3}, Al{sub 2}O{sub 3}, CaO and SiO{sub 2} by slow cooling form from a glass melt. Glass-ceramics were characterized in terms of phase assemblage, morphology, and thermal stability. The targeted phases: powellite and oxyapatite were observed in all of the compositions along with a lanthanide borosilicate, and cerianite. Results of this initial investigation of glass-ceramics show promise as a potential waste form to replace single-phase borosilicate glass.

  20. Selection of Corrosion Resistant Materials for Nuclear Waste Repositories

    SciTech Connect (OSTI)

    R.B. Rebak

    2006-08-28

    Several countries are considering geological repositories to dispose of nuclear waste. The environment of most of the currently considered repositories will be reducing in nature, except for the repository in the US, which is going to be oxidizing. For the reducing repositories, alloys such as carbon steel, stainless steels and titanium are being evaluated. For the repository in the US, some of the most corrosion resistant commercially available alloys are being investigated. This paper presents a summary of the behavior of the different materials under consideration for the repositories and the current understanding of the degradation modes of the proposed alloys in ground water environments from the point of view of general corrosion, localized corrosion and environmentally assisted cracking.

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

    SciTech Connect (OSTI)

    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 performance assessments at Hanford and across the DOE complex.

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

    SciTech Connect (OSTI)

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

    2013-04-01

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

  3. Low-level radioactive waste technology: a selected, annotated bibliography. [416 references

    SciTech Connect (OSTI)

    Fore, C.S.; Carrier, R.F.; Brewster, R.H.; Hyder, L.K.; Barnes, K.A.

    1981-10-01

    This annotated bibliography of 416 references represents the third in a series to be published by the Hazardous Materials Information Center containing scientific, technical, economic, and regulatory information relevant to low-level radioactive waste technology. The bibliography focuses on disposal site, environmental transport, and waste treatment studies as well as general reviews on the subject. The publication covers both domestic and foreign literature for the period 1951 to 1981. Major chapters selected are Chemical and Physical Aspects; Container Design and Performance; Disposal Site; Environmental Transport; General Studies and Reviews; Geology, Hydrology, and Site Resources; Regulatory and Economic Aspects; Social Aspects; Transportation Technology; Waste Production; and Waste Treatment. Entries in each of the chapters are further classified as a field study, laboratory study, theoretical study, or general overview involving one or more of these research areas.

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

    SciTech Connect (OSTI)

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

    1996-02-01

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

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

    SciTech Connect (OSTI)

    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.

  6. Selection of melter systems for the DOE/Industrial Center for Waste Vitrification Research

    SciTech Connect (OSTI)

    Bickford, D.F.

    1993-12-31

    The EPA has designated vitrification as the best developed available technology for immobilization of High-Level Nuclear Waste. In a recent federal facilities compliance agreement between the EPA, the State of Washington, and the DOE, the DOE agreed to vitrify all of the Low Level Radioactive Waste resulting from processing of High Level Radioactive Waste stored at the Hanford Site. This is expected to result in the requirement of 100 ton per day Low Level Radioactive Waste melters. Thus, there is increased need for the rapid adaptation of commercial melter equipment to DOE`s needs. DOE has needed a facility where commercial pilot scale equipment could be operated on surrogate (non-radioactive) simulations of typical DOE waste streams. The DOE/Industry Center for Vitrification Research (Center) was established in 1992 at the Clemson University Department of Environmental Systems Engineering, Clemson, SC, to address that need. This report discusses some of the characteristics of the melter types selected for installation of the Center. An overall objective of the Center has been to provide the broadest possible treatment capability with the minimum number of melter units. Thus, units have been sought which have broad potential application, and which had construction characteristics which would allow their adaptation to various waste compositions, and various operating conditions, including extreme variations in throughput, and widely differing radiological control requirements. The report discusses waste types suitable for vitrification; technical requirements for the application of vitrification to low level mixed wastes; available melters and systems; and selection of melter systems. An annotated bibliography is included.

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

    SciTech Connect (OSTI)

    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.

  8. Cadmium Chemical Form in Mine Waste Materials by X-ray Absorption Spectroscopy

    SciTech Connect (OSTI)

    Diacomanolis, V.; Ng, J. C. [University of Queensland, National Research Centre for Environmental Toxicology, Coopers Plains, QLD, 4008 (Australia); Sadler, R. [School of Public Health, Griffith University, Brisbane, QLD (Australia); Harris, H. H. [School of Chemistry and Physics, University of Adelaide, Adelaide, SA (Australia); Nomura, M. [Photon Factory, Institute of Material Structure Science, Tsukuba (Japan); Noller, B. N. [University of Queensland, Centre For Mined Land Rehabilitation, St Lucia 4072 QLD (Australia)

    2010-06-23

    This study examines the molecular form of cadmium (Cd) present in mine wastes by X-ray Absorption Spectroscopy (XAS; Cd>20 mg/kg) using the K-edge of Cd at the Photon Factory Advanced Ring (PF-AR), NW10A beam line at KEK-Tsukuba-Japan. Mine waste materials and zinc concentrate were analyzed for Cd by ICPMS prior to undertaking XAS (range 21-452 mg/kg). Model compounds (CdO, Cd(OH){sub 2}, CdCO{sub 3}, Cdacetate, CdS, Cdstearate, CdDEDTC) and samples were examined in solid form at 20 K. The XANES spectra showed similar E max values for both model compounds and samples. The EXAFS showed that Cd-S in CdS, gives a flatter spectrum in the extended region compared to Cd-O found with CdCO{sub 3}, CdO and Cd Stearate. Linear combination fitting with model Cd compounds did not give clear assignments of composition, indicating that more detailed EXAFS spectra is required as mineral forms containing Cd were present rather than simple Cd compounds such as CdCO{sub 3}. The Cd bond for a single shell model in mine waste sample matrices appears to be either Cd-O or Cd-S, or a combination of both. Comparison of molecular data from the XAS studies with bioaccessibility data giving a prediction of bioavailability for mine waste materials provides useful information about the significance of the cadmium form as a contaminant for health risk assessment purposes.

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

    SciTech Connect (OSTI)

    BLACKFORD LT

    2008-02-04

    CH2M HILL Hanford Group, Inc. (CH2M HILL) plays a critical role in Hanford Site cleanup for the U. S. Department of Energy, Office of River Protection (ORP). CH2M HILL is responsible for the management of 177 tanks containing 53 million gallons of highly radioactive wastes generated from weapons production activities from 1943 through 1990. In that time, 149 single-shell tanks, ranging in capacity from 50,000 gallons to 500,000 gallons, and 28 double-shell tanks with a capacity of 1 million gallons each, were constructed and filled with toxic liquid wastes and sludges. The cleanup mission includes removing these radioactive waste solids from the single-shell tanks to double-shell tanks for staging as feed to the Waste Treatment Plant (WTP) on the Hanford Site for vitrification of the wastes and disposal on the Hanford Site and Yucca Mountain repository. Concentrated efforts in retrieving residual solid and sludges from the single-shell tanks began in 2003; the first tank retrieved was C-106 in the 200 East Area of the site. The process for retrieval requires installation of modified sluicing systems, vacuum systems, and pumping systems into existing tank risers. Inherent with this process is the removal of existing pumps, thermo-couples, and agitating and monitoring equipment from the tank to be retrieved. Historically, these types of equipment have been extremely difficult to manage from the aspect of radiological dose, size, and weight of the equipment, as well as their attendant operating and support systems such as electrical distribution and control panels, filter systems, and mobile retrieval systems. Significant effort and expense were required to manage this new waste stream and resulted in several events over time that were both determined to be unsafe for workers and potentially unsound for protection of the environment. Over the last four years, processes and systems have been developed that reduce worker exposures to these hazards, eliminate violations 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.

  10. Los Alamos National Laboratory selects small businesses for nuclear waste

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would likeUniverseIMPACTThousand CubicResourcelogo and-Eearns threeservices LANL selects

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

    DOE Patents [OSTI]

    Colombo, P.; Kalb, P.D.

    1984-06-05

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

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

    SciTech Connect (OSTI)

    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, 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 as durable as LAW glass. Monolithing of the granular FBSR product, which is one of the objectives of this current study, is being investigated to prevent dispersion during transport or burial/storage but is not necessary for performance. FBSR testing of a Hanford LAW simulant and a WTP-SW simulant at the pilot scale was performed by THOR Treatment Technologies, LLC at Hazen Research Inc. in April/May 2008. The Hanford LAW simulant was the Rassat 68 tank blend and the target concentrations for the LAW was increased by a factor of 10 for Sb, As, Ag, Cd, and Tl; 100 for Ba and Re (Tc surrogate); 1,000 for I; and 254,902 for Cs based on discussions with the DOE field office and the environmental regulators and an evaluation of the Hanford Tank Waste Envelopes A, B, and C. It was determined through the evaluation of the actual tank waste metals concentrations that some metal levels were not sufficient to achieve reliable detection in the off-gas sampling. Therefore, the identified metals concentrations were increased in the Rassat simulant processed by TTT at HRI to ensure detection and enable calculation of system removal efficiencies, product retention efficiencies, and mass balance closure without regard to potential results of those determinations or impacts on product durability response such as Toxicity Characteristic Leach Procedure (TCLP). A WTP-SW simulant based on melter off-gas analyses from Vitreous State Laboratory (VSL) was also tested at HRI in the 15-inch diameter Engineering Scale Test Demonstration (ESTD) dual reformer at HRI in 2008. The target concentrations for the Resource Conservation and Recovery Act (RCRA) metals were increased by 16X for Se, 29X for Tl, 42X for Ba, 48X for Sb, by 100X for Pb and Ni, 1000X for Ag, and 1297X for Cd to ensure detection by the an

  13. Global Nuclear Energy Partnership Waste Treatment Baseline

    SciTech Connect (OSTI)

    Dirk Gombert; William Ebert; James Marra; Robert Jubin; John Vienna

    2008-05-01

    The Global Nuclear Energy Partnership program (GNEP) is designed to demonstrate a proliferation-resistant and sustainable integrated nuclear fuel cycle that can be commercialized and used internationally. Alternative stabilization concepts for byproducts and waste streams generated by fuel recycling processes were evaluated and a baseline of waste forms was recommended for the safe disposition of waste streams. Waste forms are recommended based on the demonstrated or expected commercial practicability and technical maturity of the processes needed to make the waste forms, and performance of the waste form materials when disposed. Significant issues remain in developing technologies to process some of the wastes into the recommended waste forms, and a detailed analysis of technology readiness and availability may lead to the choice of a different waste form than what is recommended herein. Evolving regulations could also affect the selection of waste forms.

  14. Volatilization of selected organic compounds from a creosote-waste land-treatment facility. Master's thesis

    SciTech Connect (OSTI)

    Scott, E.J.

    1989-01-01

    The purpose of this research was to evaluate the emissions of volatile and semi-volatile compounds which are constituents of a complex creosote waste from laboratory simulations of a land treatment system to assess the potential human exposure to hazardous compounds from this source. In addition, the Thibodeaux-Hwang Air Emission Release Rate (AERR) model was evaluated for its use in predicting emission rates of hazardous constituents of creosote wood preservative waste from land treatment facilities. A group of hazardous volatile and semi-volatile constituents present in the creosote waste was selected for evaluation in this study and included a variety of polynuclear aromatic hydrocarbons (PNA's), phenol, and chlorinated and substituted phenols.

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

    SciTech Connect (OSTI)

    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.

  16. Annual Benefits Enrollment Form 2012 Plan Year Select Campus Location: Norman Oklahoma City Tulsa

    E-Print Network [OSTI]

    Oklahoma, University of

    Annual Benefits Enrollment Form 2012 Plan Year Select Campus Location: Norman Oklahoma City Tulsa BlueLincs HMO HMO Primary Care Physician #: Community Care HMO (Tulsa Area Only) HMO Primary Care

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

    SciTech Connect (OSTI)

    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.

  18. Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model - 13413

    SciTech Connect (OSTI)

    Djokic, Denia [Department of Nuclear Engineering, University of California - Berkeley, 4149 Etcheverry Hall, Berkeley, CA 94720-1730 (United States)] [Department of Nuclear Engineering, University of California - Berkeley, 4149 Etcheverry Hall, Berkeley, CA 94720-1730 (United States); Piet, Steven J.; Pincock, Layne F.; Soelberg, Nick R. [Idaho National Laboratory - INL, 2525 North Fremont Avenue, Idaho Falls, ID 83415 (United States)] [Idaho National Laboratory - INL, 2525 North Fremont Avenue, Idaho Falls, ID 83415 (United States)

    2013-07-01

    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. (authors)

  19. DOI: 10.1002/adma.200500769 Unique Properties of Selectively Formed Zirconia Nanostructures

    E-Print Network [OSTI]

    Wang, Zhong L.

    DOI: 10.1002/adma.200500769 Unique Properties of Selectively Formed Zirconia Nanostructures By James L. Gole,* Sharka M. Prokes, John D. Stout, Orest J. Glembocki, and Rusen Yang Zirconia has forms, but also on the first synthesis procedure for preparing zirconia-based nanoshells and hollow

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

    SciTech Connect (OSTI)

    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.

  1. Fuzzy multicriteria disposal method and site selection for municipal solid waste

    SciTech Connect (OSTI)

    Ekmekcioglu, Mehmet; Kaya, Tolga; Kahraman, Cengiz

    2010-08-15

    The use of fuzzy multiple criteria analysis (MCA) in solid waste management has the advantage of rendering subjective and implicit decision making more objective and analytical, with its ability to accommodate both quantitative and qualitative data. In this paper a modified fuzzy TOPSIS methodology is proposed for the selection of appropriate disposal method and site for municipal solid waste (MSW). Our method is superior to existing methods since it has capability of representing vague qualitative data and presenting all possible results with different degrees of membership. In the first stage of the proposed methodology, a set of criteria of cost, reliability, feasibility, pollution and emission levels, waste and energy recovery is optimized to determine the best MSW disposal method. Landfilling, composting, conventional incineration, and refuse-derived fuel (RDF) combustion are the alternatives considered. The weights of the selection criteria are determined by fuzzy pairwise comparison matrices of Analytic Hierarchy Process (AHP). It is found that RDF combustion is the best disposal method alternative for Istanbul. In the second stage, the same methodology is used to determine the optimum RDF combustion plant location using adjacent land use, climate, road access and cost as the criteria. The results of this study illustrate the importance of the weights on the various factors in deciding the optimized location, with the best site located in Catalca. A sensitivity analysis is also conducted to monitor how sensitive our model is to changes in the various criteria weights.

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

    SciTech Connect (OSTI)

    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.

  3. Ion exchange columns for selective removal of cesium from aqueous radioactive waste using hydrous crystalline silico-titanates 

    E-Print Network [OSTI]

    Ricci, David Michael

    1995-01-01

    conscious society. In Hanford, WA, hundreds of underground storage tanks hold tens of millions of gallons of aqueous radioactive waste. This liquid waste, which has a very high sodium content, contains trace amounts of radioactive cesium 137. Since.... The radioactive waste would be concentrated in the exchanger so that it can be stored in solid form, which requires less storage space. The remaining liquid would no longer be radioactive and could be disposed of at a much lower cost. Ion exchange has...

  4. Determination of hexabromocyclododecane in waste waters with a bromide-selective electrode

    SciTech Connect (OSTI)

    Paama, L.A.; Kokk, K.Y.

    1985-11-01

    This paper presents a method for the determination of microquantities of hexabromocyclodecane (HBCD), an antipyrene for the reduction of the flammability of frothing polystyrene in industrial waste waters. The method is based on the extraction of HBCD from waste waters with chloroform, evaporation of the chloroform from the extracts in a stream of air, mineralization of the dry residues, and determination of the bromide ions with a bromine-selective electrode. One of the simpler approaches is given as based on alkaline hydrolysis with dimethyl formamide or in the presence of metallic cadmium. The dry residues after evaporation of the chloroform were decomposed by boiling in a KOH solution in the presence of metallic zinc.

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

    SciTech Connect (OSTI)

    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(am,hyd)} is also likely to be present in deposits and scales that have formed on the steel surfaces of the tank. Over the operational period and after closure of Tank 18, Ostwald ripening has and will continue to transform PuO{sub 2(am,hyd)} to a more crystalline form of plutonium dioxide, PuO{sub 2(c)}. After bulk waste removal and heel retrieval operations, the free hydroxide concentration decreased and the carbonate concentration in the free liquid and solids increased. Consequently, a portion of the PuO{sub 2(am,hyd)} has likely been converted to a hydroxy-carbonate complex such as Pu(OH){sub 2}(CO{sub 3}){sub (s)}. or PuO(CO{sub 3}) {center_dot} xH{sub 2}O{sub (am)}. Like PuO{sub 2(am,hyd)}, Ostwald ripening of Pu(OH){sub 2}(CO{sub 3}){sub (s)} or PuO(CO{sub 3}) {center_dot} xH{sub 2}O{sub (am)} would be expected to occur to produce a more crystalline form of the plutonium carbonate complex. Due to the high alkalinity and low carbonate concentration in the grout formulation, it is expected that upon interaction with the grout, the plutonium carbonate complexes will transform back into plutonium hydroxide. Although crystalline plutonium dioxide is the more stable thermodynamic state of Pu(IV), the low temperature and high water content of the waste during the operating and heel removal periods in Tank 18 have limited the transformation of the plutonium into crystalline plutonium dioxide. During the tank closure period of thousands of years, transformation of the plutonium into a more crystalline plutonium dioxide form would be expected. However, the continuing presence of water, reaction with water radiolysis products, and low temperatures will limit the transformation, and will likely maintain an amorphous Pu(OH){sub 4} or PuO{sub 2(am,hyd)} form on the surface of any crystalline plutonium dioxide produced after tank closure. X-ray Absorption Spectroscopic (XAS) measurements of Tank 18 residues are recommended to confirm coordination environments of the plutonium. If the presence of PuO(CO{sub 3}){sub (am,hyd)} is confirmed by XAS, it is recommended that e

  6. Characterization and Leaching Tests of the Fluidized Bed Steam Reforming (FBSR) Waste Form for LAW Immobilization

    SciTech Connect (OSTI)

    Neeway, James J.; Qafoku, Nikolla; Brown, Christopher F.; Peterson, Reid A.

    2013-10-01

    Several supplemental technologies for treating and immobilizing Hanford low activity waste (LAW) have been evaluated. One such immobilization technology is the Fluidized Bed Steam Reforming (FBSR) granular product. The FBSR granular product is composed of insoluble sodium aluminosilicate (NAS) feldspathoid minerals. Production of the FBSR mineral product has been demonstrated both at the industrial and laboratory scale. Pacific Northwest National Laboratory (PNNL) was involved in an extensive characterization campaign. This goal of this campaign was study the durability of the FBSR mineral product and the mineral product encapsulated in a monolith to meet compressive strength requirements. This paper gives an overview of results obtained using the ASTM C 1285 Product Consistency Test (PCT), the EPA Test Method 1311 Toxicity Characteristic Leaching Procedure (TCLP), and the ASTMC 1662 Single-Pass Flow-Through (SPFT) test. Along with these durability tests an overview of the characteristics of the waste form has been collected using Scanning Electron Microscopy (SEM), X-ray Diffraction (XRD), microwave digestions for chemical composition, and surface area from Brunauer, Emmett, and Teller (BET) theory.

  7. Cement waste-form development for ion-exchange resins at the Rocky Flats Plant

    SciTech Connect (OSTI)

    Veazey, G.W.; Ames, R.L.

    1997-03-01

    This report describes the development of a cement waste form to stabilize ion-exchange resins at Rocky Flats Environmental Technology Site (RFETS). These resins have an elevated potential for ignition due to inadequate wetness and contact with nitrates. The work focused on the preparation and performance evaluation of several Portland cement/resin formulations. The performance standards were chosen to address Waste Isolation Pilot Plant and Environmental Protection Agency Resource Conservation and Recovery Act requirements, compatibility with Rocky Flats equipment, and throughput efficiency. The work was performed with surrogate gel-type Dowex cation- and anion-exchange resins chosen to be representative of the resin inventory at RFETS. Work was initiated with nonactinide resins to establish formulation ranges that would meet performance standards. Results were then verified and refined with actinide-containing resins. The final recommended formulation that passed all performance standards was determined to be a cement/water/resin (C/W/R) wt % ratio of 63/27/10 at a pH of 9 to 12. The recommendations include the acceptable compositional ranges for each component of the C/W/R ratio. Also included in this report are a recommended procedure, an equipment list, and observations/suggestions for implementation at RFETS. In addition, information is included that explains why denitration of the resin is unnecessary for stabilizing its ignitability potential.

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

    SciTech Connect (OSTI)

    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.

  9. Commercial waste treatment program annual progress report for FY 1983

    SciTech Connect (OSTI)

    McElroy, J.L.; Burkholder, H.C. (comps.)

    1984-02-01

    This annual report describes progress during FY 1983 relating to technologies under development by the Commercial Waste Treatment Program, including: development of glass waste form and vitrification equipment for high-level wastes (HLW); waste form development and process selection for transuranic (TRU) wastes; pilot-scale operation of a radioactive liquid-fed ceramic melter (LFCM) system for verifying the reliability of the reference HLW treatment proces technology; evaluation of treatment requirements for spent fuel as a waste form; second-generation waste form development for HLW; and vitrification process control and product quality assurance technologies.

  10. ACCOUNTING FOR A VITRIFIED PLUTONIUM WASTE FORM IN THE YUCCA MOUNTAIN REPOSITORY TOTAL SYSTEM PERFORMANCE ASSESSMENT (TSPA)

    SciTech Connect (OSTI)

    Marra, J

    2007-02-12

    A vitrification technology utilizing a lanthanide borosilicate (LaBS) glass appears to be a viable option for dispositioning excess weapons-useable plutonium that is not suitable for processing into mixed oxide (MOX) fuel. A significant effort to develop a glass formulation and vitrification process to immobilize plutonium was completed in the mid-1990s to support the Plutonium Immobilization Program (PIP). Further refinement of the vitrification process was accomplished as part of the Am/Cm solution vitrification project. The LaBS glass formulation was found to be capable of immobilizing in excess of 10 wt% Pu and to be very tolerant of the impurities accompanying the plutonium material streams. Thus, this waste form would be suitable for dispositioning plutonium owned by the Department of Energy-Office of Environmental Management (DOE-EM) that may not be well characterized and may contain high levels of impurities. The can-in-canister technology demonstrated in the PIP could be utilized to dispose of the vitrified plutonium in the federal radioactive waste repository. The can-in-canister technology involves placing small cans of the immobilized Pu form into a high level waste (HLW) glass canister fitted with a rack to hold the cans and then filling the canister with HLW glass. Testing was completed to demonstrate that this technology could be successfully employed with little or no impact to current Defense Waste Processing Facility (DWPF) operation and that the resulting canisters were essentially equivalent to the present HLW glass canisters to be dispositioned in the federal repository. The performance of wastes in the repository and, moreover, the performance of the entire repository system is being evaluated by the Department of Energy-Office of Civilian Radioactive Waste Management (DOE-RW) using a Total System Performance Assessment (TSPA) methodology. Technical bases documents (e.g., Analysis/Modeling Reports (AMR)) that address specific issues regarding waste form performance are being used to develop process models as input to the TSPA analyses. In this report, models developed in five AMRs for waste forms currently slated for disposition in the repository are evaluated for their applicability to waste forms with plutonium immobilized in LaBS glass using the can-in-canister technology. Those AMRs address: high-level waste glass degradation; radionuclide inventory; in-package chemistry; dissolved concentration limits of radioactive elements; and colloid-associated radionuclide concentrations. Based on evaluation of how the models treated HLW glass and similarities in the corrosion behaviors of borosilicate HLW glasses and LaBS glass, the models in the AMRs were deemed to be directly applicable to the disposition of excess weapons-useable plutonium. The evaluations are summarized.

  11. Selected, annotated bibliography of studies relevant to the isolation of nuclear wastes. [705 references

    SciTech Connect (OSTI)

    Hyder, L.K.; Fore, C.S.; Vaughan, N.D.; Faust, R.A.

    1980-09-01

    This annotated bibliography of 705 references represents the first in a series to be published by the Ecological Sciences Information Center containing scientific, technical, economic, and regulatory information relevant to nuclear waste isolation. Most references discuss deep geologic disposal, with fewer studies of deep seabed disposal; space disposal is also included. The publication covers both domestic and foreign literature for the period 1954 to 1980. Major chapters selected are Chemical and Physical Aspects; Container Design and Performance; Disposal Site; Envirnmental Transport; General Studies and Reviews; Geology, Hydrology and Site Resources; Regulatory and Economic Aspects; Repository Design and Engineering; Transportation Technology; Waste Production; and Waste Treatment. Specialized data fields have been incorporated to improve the ease and accuracy of locating pertinent references. Specific radionuclides for which data are presented are listed in the Measured Radionuclides field, and specific parameters which affect the migration of these radionuclides are presented in the Measured Parameters field. The references within each chapter are arranged alphabetically by leading author, corporate affiliation, or title of the document. When the author is not given, the corporate affiliation appears first. If these two levels of authorship are not given, the title of the document is used as the identifying level. Indexes are provided for author(s), keywords, subject category, title, geographic location, measured parameters, measured radionuclides, and publication description.

  12. Comparison of selected foreign plans and practices for spent fuel and high-level waste management

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.; Johnson, A.B. Jr.; Hazelton, R.F.; Bradley, D.J.

    1990-04-01

    This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.

  13. Comparison of low-level waste disposal programs of DOE and selected international countries

    SciTech Connect (OSTI)

    Meagher, B.G. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); Cole, L.T. [Cole and Associates (United States)

    1996-06-01

    The purpose of this report is to examine and compare the approaches and practices of selected countries for disposal of low-level radioactive waste (LLW) with those of the US Department of Energy (DOE). The report addresses the programs for disposing of wastes into engineered LLW disposal facilities and is not intended to address in-situ options and practices associated with environmental restoration activities or the management of mill tailings and mixed LLW. The countries chosen for comparison are France, Sweden, Canada, and the United Kingdom. The countries were selected as typical examples of the LLW programs which have evolved under differing technical constraints, regulatory requirements, and political/social systems. France was the first country to demonstrate use of engineered structure-type disposal facilities. The UK has been actively disposing of LLW since 1959. Sweden has been disposing of LLW since 1983 in an intermediate-depth disposal facility rather than a near-surface disposal facility. To date, Canada has been storing its LLW but will soon begin operation of Canada`s first demonstration LLW disposal facility.

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

    SciTech Connect (OSTI)

    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 isotopes I and Re as analogs to {sup 129}I and {sup 99}Tc. The results have confirmed our original finding that nanoporous Al oxide and its derivatives have high I sorption capabilities due to the combined effects of surface chemistry and nanopore confinement. We have developed a suite of techniques for the fixation of radionuclides in metal oxide nanopores. The key to this fixation is to chemically convert a target radionuclide into a less volatile or soluble form. We have developed a technique to convert a radionuclide-loaded nanoporous material into a durable glass-ceramic waste form through calcination. We have shown that mixing a radionuclide-loaded getter material with a Na-silicate solution can effectively seal the nanopores in the material, thus enhancing radionuclide retention during waste form formation. Our leaching tests have demonstrated the existence of an optimal vitrification temperature for the enhancement of waste form durability. Our work also indicates that silver may not be needed for I immobilization and encapsulation.

  15. Crystalline Ceramic Waste Forms: Report Detailing Data Collection In Support Of Potential FY13 Pilot Scale Melter Test

    SciTech Connect (OSTI)

    Brinkman, K. S.; Amoroso, J.; Marra, J. C.; Fox, K. M.

    2012-09-21

    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 summarize the data collection in support of future melter demonstration testing for crystalline ceramic waste forms. 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. The principal difficulties encountered during processing of the ?reference ceramic? waste form by a melt and crystallization process were the incomplete incorporation of Cs into the hollandite phase and the presence of secondary Cs-Mo non-durable phases. In the single phase hollandite system, these issues were addressed in this study by refining the compositions to include Cr as a transition metal element and the use of Ti/TiO{sub 2} buffer to maintain reducing conditions. Initial viscosity studies of ceramic waste forms indicated that the pour spout must be maintained above 1400{deg}C to avoid flow blockages due to crystallization. In-situ electron irradiations simulate radiolysis effects indicated hollandite undergoes a crystalline to amorphous transition after a radiation dose of 10{sup 13} Gy which corresponds to approximately 1000 years at anticipated doses (2?10{sup 10}-2?10{sup 11} Gy). Dual-beam ion irradiations employing light ion beam (such as 5 MeV alpha) and heavy ion beam (such as 100 keV Kr) studies indicate that reference ceramic waste forms are radiation tolerant to the ??particles and ?-particles, but are susceptible to a crystalline to amorphous transition under recoil nuclei effects. A path forward for refining the processing steps needed to form the targeted phase assemblages is outlined in this report. Processing modifications including melting in a reducing atmosphere with the use of Ti/TiO2 buffers, and the addition of Cr to the transition metal additives to facilitate Cs-incorporation in the hollandite phase. In addition to melt processing, alternative fabrication routes are being considered including Spark Plasma Sintering (SPS) and Hot Isostatic Pressing (HIP).

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

    SciTech Connect (OSTI)

    Taga, Raijeli L.; Ng, Jack [University of Queensland, National Research Centre for Environmental Toxicology (EnTox), Brisbane, 4108 (Australia); Zheng Jiajia; Huynh, Trang; Noller, Barry [University of Queensland, Centre for Mined Land Rehabilitation, Brisbane, 4072 (Australia); Harris, Hugh H. [School of Chemistry and Physics, University of Adelaide, Adelaide, 5005 (Australia)

    2010-06-23

    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 (PbFe{sub 6}(SO{sub 4}){sub 4}(OH){sub 12}) by lead L{sub 3}-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 PM{sub 10} 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.

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

    SciTech Connect (OSTI)

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

    1999-10-25

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

  18. Milestones for Selection, Characterization, and Analysis of the Performance of a Repository for Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain.

    SciTech Connect (OSTI)

    Rechard, Robert P.

    2014-02-01

    This report presents a concise history in tabular form of events leading up to site identification in 1978, site selection in 1987, subsequent characterization, and ongoing analysis through 2008 of the performance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain in southern Nevada. The tabulated events generally occurred in five periods: (1) commitment to mined geologic disposal and identification of sites; (2) site selection and analysis, based on regional geologic characterization through literature and analogous data; (3) feasibility analysis demonstrating calculation procedures and importance of system components, based on rough measures of performance using surface exploration, waste process knowledge, and general laboratory experiments; (4) suitability analysis demonstrating viability of disposal system, based on environment-specific laboratory experiments, in-situ experiments, and underground disposal system characterization; and (5) compliance analysis, based on completed site-specific characterization. Because the relationship is important to understanding the evolution of the Yucca Mountain Project, the tabulation also shows the interaction between four broad categories of political bodies and government agencies/institutions: (a) technical milestones of the implementing institutions, (b) development of the regulatory requirements and related federal policy in laws and court decisions, (c) Presidential and agency directives and decisions, and (d) critiques of the Yucca Mountain Project and pertinent national and world events related to nuclear energy and radioactive waste.

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

    SciTech Connect (OSTI)

    Jantzen, C. M.; Crawford, C. L.; Bannochie, C. J.; Burket, P. R.; Cozzi, A. D.; Daniel, W. E.; Hall, H. K.; Miller, D. H.; Missimer, D. M.; Nash, C. A.; Williams, M. F.

    2013-08-21

    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 amorphous, macro-encapsulates the granules, and the monoliths pass ANSI/ANS 16.1 and ASTM C1308 durability testing with Re achieving a Leach Index (LI) of 9 (the Hanford Integrated Disposal Facility, IDF, criteria for Tc-99) after a few days and Na achieving an LI of >6 (the Hanford IDF criteria for Na) in the first few hours. The granular and monolithic waste forms also pass the EPA Toxicity Characteristic Leaching Procedure (TCLP) for all Resource Conservation and Recovery Act (RCRA) components at the Universal Treatment Standards (UTS). Two identical Benchscale Steam Reformers (BSR) were designed and constructed at SRNL, one to treat non-radioactive simulants and the other to treat actual radioactive wastes. The results from the non-radioactive BSR were used to determine the parameters needed to operate the radioactive BSR in order to confirm the findings of non-radioactive FBSR pilot scale and engineering scale tests and to qualify an FBSR LAW waste form for applications at Hanford. Radioactive testing commenced using SRS LAW from Tank 50 chemically trimmed to look like Hanford’s blended LAW known as the Rassat simulant as this simulant composition had been tested in the non-radioactive BSR, the non-radioactive pilot scale FBSR at the Science Applications International Corporation-Science and Technology Applications Research (SAIC-STAR) facility in Idaho Falls, ID and in the TTT Engineering Scale Technology Demonstration (ESTD) at Hazen Research Inc. (HRI) in Denver, CO. This provided a “tie back” between radioactive BSR testing and non-radioactive BSR, pilot scale, and engineering scale testing. Approximately six hundred grams of non-radioactive and radioactive BSR product were made for extensive testing and comparison to the non-radioactive pilot scale tests performed in 2004 at SAIC-STAR and the engineering scale test performed in 2008 at HRI with the Rassat simulant. The same mineral phases and off-gas species were found in the radioactive and non-radioactive testing. The granular ESTD and BSR products (radioactive and non-radioactive) were analyzed for to

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  2. A leach model for solidified/stabilized waste forms based on empirical partitioning of contaminants 

    E-Print Network [OSTI]

    Kim, Inchul

    1997-01-01

    An Empirical Partitioning Leach Model (EPLEM), has been developed to describe the leaching of metal contaminants from solidified wastes. The model considers interactions between two components such as hydrogen ion and a metal. The hydrogen ion...

  3. Subterranean barriers, methods, and apparatuses for forming, inspecting, selectively heating, and repairing same

    DOE Patents [OSTI]

    Nickelson, Reva A. (Shelley, ID); Sloan, Paul A. (Rigby, ID); Richardson, John G. (Idaho Falls, ID); Walsh, Stephanie (Idaho Falls, ID); Kostelnik, Kevin M. (Idaho, ID)

    2009-04-07

    A subterranean barrier and method for forming same are disclosed, the barrier including a plurality of casing strings wherein at least one casing string of the plurality of casing strings may be affixed to at least another adjacent casing string of the plurality of casing strings through at least one weld, at least one adhesive joint, or both. A method and system for nondestructively inspecting a subterranean barrier is disclosed. For instance, a radiographic signal may be emitted from within a casing string toward an adjacent casing string and the radiographic signal may be detected from within the adjacent casing string. A method of repairing a barrier including removing at least a portion of a casing string and welding a repair element within the casing string is disclosed. A method of selectively heating at least one casing string forming at least a portion of a subterranean barrier is disclosed.

  4. Alternatives Generation and Analysis for Phase 1 High Level Waste Feed Tanks Selection

    SciTech Connect (OSTI)

    CRAWFORD, T.W.

    1999-08-16

    A recent revision of the US. Department of Energy privatization contract for the immobilization of high-level waste (HLW) at Hanford necessitates the investigation of alternative waste feed sources to meet contractual feed requirements. This analysis identifies wastes to be considered as HLW feeds and develops and conducts alternative analyses to comply with established criteria. A total of 12,426 cases involving 72 waste streams are evaluated and ranked in three cost-based alternative models. Additional programmatic criteria are assessed against leading alternative options to yield an optimum blended waste feed stream.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  7. Use of Novel Highly Selective Ion Exchange Media for Minimizing the Waste Arising from Different NPP and Other Liquids

    SciTech Connect (OSTI)

    Tusa, Esko; Harjula, Risto; Lehto, Jukka

    2003-02-25

    Highly selective inorganic ion exchangers give new possibilities to implement and operate new innovative treatment systems for radioactive liquids. Because of high selectivity these ion exchangers can be used even in liquids of high salt concentrations. Only selected target nuclides will be separated and inactive salts are left in the liquid, which can be released or recategorized. Thus, it is possible to reduce the volume of radioactive waste dramatically. On the other hand, only a small volume of highly selective material is required in applications, which makes it possible to design totally new types of compact treatment systems. The major benefit of selective ion exchange media comes from the very large volume reduction of radioactive waste in final disposal. It is also possible to save in investment costs, because small ion exchanger volumes can be used and handled in a very small facility. This paper describes different applications of these highly selective ion exchangers, both commercial fullscale applications and laboratory tests, to give the idea of their efficiency for different liquids.

  8. COMPACTING BIOMASS AND MUNICIPAL SOLID WASTES TO FORM AND UPGRADED FUEL

    SciTech Connect (OSTI)

    Henry Liu; Yadong Li

    2000-11-01

    Biomass waste materials exist in large quantity in every city and in numerous industrial plants such as wood processing plants and waste paper collection centers. Through minimum processing, such waste materials can be turned into a solid fuel for combustion at existing coal-fired power plants. Use of such biomass fuel reduces the amount of coal used, and hence reduces the greenhouse effect and global warming, while at the same time it reduces the use of land for landfill and the associated problems. The carbon-dioxide resulting from burning biomass fuel is recycled through plant growth and hence does not contribute to global warming. Biomass fuel also contains little sulfur and hence does not contribute to acid rain problems. Notwithstanding the environmental desirability of using biomass waste materials, not much of them are used currently due to the need to densify the waste materials and the high cost of conventional methods of densification such as pelletizing and briquetting. The purpose of this project was to test a unique new method of biomass densification developed from recent research in coal log pipeline (CLP). The new method can produce large agglomerates of biomass materials called ''biomass logs'' which are more than 100 times larger and 30% denser than conventional ''pellets'' or ''briquettes''. The Phase I project was to perform extensive laboratory tests and an economic analysis to determine the technical and economic feasibility of the biomass log fuel (BLF). A variety of biomass waste materials, including wood processing residues such as sawdust, mulch and chips of various types of wood, combustibles that are found in municipal solid waste stream such as paper, plastics and textiles, energy crops including willows and switch grass, and yard waste including tree trimmings, fallen leaves, and lawn grass, were tested by using this new compaction technology developed at Capsule Pipeline Research Center (CPRC), University of Missouri-Columbia (MU). The compaction conditions, including compaction pressure, pressure holding time, back pressure, moisture content, particle size and shape, piston and mold geometry and roughness, and binder for the materials were studied and optimized. The properties of the compacted products--biomass logs--were evaluated in terms of physical, mechanical, and combustion characteristics. An economic analysis of this technology for anticipated future commercial operations was performed. It was found that the compaction pressure and the moisture content of the biomass materials are critical for producing high-quality biomass logs. For most biomass materials, dense and strong logs can be produced under room temperature without binder and at a pressure of 70 MPa (10,000 psi), approximately. A few types of the materials tested such as sawdust and grass need a minimum pressure of 100 MPa (15,000 psi) in order to produce good logs. The appropriate moisture range for compacting waste paper into good logs is 5-20%, and the optimum moisture is in the neighborhood of 13%. For the woody materials and yard waste, the appropriate moisture range is narrower: 5-13%, and the optimum is 8-9%. The compacted logs have a dry density of 0.8 to 1.0 g/cm{sup 3}, corresponding to a wet density of 0.9 to 1.1 g/cm{sup 3}, approximately. The logs have high strength and high resistance to impact and abrasion, but are feeble to water and hence need to be protected from water or rain. They also have good long-term performance under normal environmental conditions, and can be stored for a long time without significant deterioration. Such high-density and high-strength logs not only facilitate handling, transportation, and storage, but also increase the energy content of biomass per unit volume. After being transported to power plants and crushed, the biomass logs can be co-fired with coal to generate electricity.

  9. Wastes from selected activities in two light-water reactor fuel cycles

    SciTech Connect (OSTI)

    Palmer, C.R.; Hill, O.F.

    1980-07-01

    This report presents projected volumes and radioactivities of wastes from the production of electrical energy using light-water reactors (LWR). The projections are based upon data developed for a recent environmental impact statement in which the transuranic wastes (i.e., those wastes containing certain long-lived alpha emitters at concentrations of at least 370 becquerels, or 10 nCi, per gram of waste) from fuel cycle activities were characterized. In addition, since the WG.7 assumed that all fuel cycle wastes except mill tailings are placed in a mined geologic repository, the nontransuranic wastes from several activities are included in the projections reported. The LWR fuel cycles considered are the LWR, once-through fuel cycle (Strategy 1), in which spent fuel is packaged in metal canisters and then isolated in geologic formations; and the LWR U/Pu recycle fuel cycle (Strategy 2), wherein spent fuel is reprocessed for recovery and recycle of uranium and plutonium in LWRs. The wastes projected for the two LWR fuel cycles are summarized. The reactor operations and decommissioning were found to dominate the rate of waste generation in each cycle. These activities account for at least 85% of the fuel cycle waste volume (not including head-end wastes) when normalized to per unit electrical energy generated. At 10 years out of reactor, however, spent fuel elements in Strategy 1 represent 98% of the fuel cycle activity but only 4% of the volume. Similarly, the packaged high-level waste, fuel hulls and hardware in Strategy 2 concentrate greater than 95% of the activity in 2% of the waste volume.

  10. RADIOACTIVE DEMONSTRATION OF FINAL MINERALIZED WASTE FORMS FOR HANFORD WASTE TREATMENT PLANT SECONDARY WASTE (WTP-SW) BY FLUIDIZED BED STEAM REFORMING (FBSR) USING THE BENCH SCALE REFORMER PLATFORM

    SciTech Connect (OSTI)

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

    2014-08-21

    The U.S. Department of Energy’s Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford’s tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as {sup 137}Cs, {sup 129}I, {sup 99}Tc, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150°C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750°C) continuous method by which WTP-SW wastes can be processed irrespective of whether they 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 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. The granular products (both simulant and radioactive) were tested and a subset of the granular material (both simulant and radioactive) were stabilized in a geopolymer matrix. Extensive testing and characterization of the granular and monolith material were made including the following: ? ASTM C1285 (Product Consistency Test) testing of granular and monolith; ? ASTM C1308 accelerated leach testing of the radioactive monolith; ? ASTM C192 compression testing of monoliths; and ? EPA Method 1311 Toxicity Characteristic Leaching Procedure (TCLP) testing. The significant findings of the testing completed on simulant and radioactive WTP-SW are given below: ? Data indicates {sup 99}Tc, Re, Cs, and I

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

    SciTech Connect (OSTI)

    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.

  12. Evaluation of selected detector systems for products formed in the atmospheric hydrolysis of uranium hexafluoride

    SciTech Connect (OSTI)

    Bostick, W.D.; Bostick, D.T.

    1987-03-01

    Sensitive detection of UF/sub 6/ hydrolysis products, either by discontinuous sampling or by continuous or near real-time monitoring, is an important safety consideration for DOE contractors handling large quantities of UF/sub 6/. Automated continuous or rapid intermittent remote sensing of these reaction products can provide an alarm signal when a preselected threshold value has been exceeded (absolute response) or when a significant emission excursion has occurred (rate of change of response). This report evaluates the performance of selected devices for the detection of airborne materials formed in the release of liquid UF/sub 6/ (approx. =1.3 g) into an enclosed volume of 6 m/sup 3/; these experiments were initiated on October 23, 1986. The detection principles investigated are: photometric, gas detector tubes, and electrochemical sensor.

  13. Low Temperature Waste Immobilization Testing Vol. I

    SciTech Connect (OSTI)

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

    2006-09-14

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

  14. Demonstration and Transfer of Selected New Technologies for Animal Waste Pollution Control 

    E-Print Network [OSTI]

    Mukhtar, Saqib; Gregory, Lucas

    2009-01-01

    ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 0 Waste w a t e r Treat me n t Solut i o n (WTS ? ) System ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 Sa mple and Sa mple Analysis... ................................................................................................15 L4DB ? Microbia l Treatmen t System ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 5 Waste w a t e r Treat me n t Solut i o n (WTS...

  15. Web-GIS oriented systems viability for municipal solid waste selective collection optimization in developed and transient economies

    SciTech Connect (OSTI)

    Rada, E.C.; Ragazzi, M.; Fedrizzi, P.

    2013-04-15

    Highlights: ? As an appropriate solution for MSW management in developed and transient countries. ? As an option to increase the efficiency of MSW selective collection. ? As an opportunity to integrate MSW management needs and services inventories. ? As a tool to develop Urban Mining actions. - Abstract: Municipal solid waste management is a multidisciplinary activity that includes generation, source separation, storage, collection, transfer and transport, processing and recovery, and, last but not least, disposal. The optimization of waste collection, through source separation, is compulsory where a landfill based management must be overcome. In this paper, a few aspects related to the implementation of a Web-GIS based system are analyzed. This approach is critically analyzed referring to the experience of two Italian case studies and two additional extra-European case studies. The first case is one of the best examples of selective collection optimization in Italy. The obtained efficiency is very high: 80% of waste is source separated for recycling purposes. In the second reference case, the local administration is going to be faced with the optimization of waste collection through Web-GIS oriented technologies for the first time. The starting scenario is far from an optimized management of municipal solid waste. The last two case studies concern pilot experiences in China and Malaysia. Each step of the Web-GIS oriented strategy is comparatively discussed referring to typical scenarios of developed and transient economies. The main result is that transient economies are ready to move toward Web oriented tools for MSW management, but this opportunity is not yet well exploited in the sector.

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

    SciTech Connect (OSTI)

    Grutzeck, Michael W.

    2005-06-27

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

  17. 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

    E-Print Network [OSTI]

    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...

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

    SciTech Connect (OSTI)

    Alaniz, Ariana J.; Delgado, Luc R.; Werbick, Brett M.; Hartmann, Thomas

    2013-07-01

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

  19. Design, optimization, and selectivity of inorganic ion-exchangers for radioactive waste remediation 

    E-Print Network [OSTI]

    Medvedev, Dmitry Gennadievich

    2005-11-01

    of the clean up effort is targeted towards High Level Waste (HLW), with 94 million gallons of untreated HLW in storage at DOE facilities at the Hanford Site in Washington, the Savannah River Site (SRS) in South Carolina, and the Idaho National Engineering... complicated after September 11 of 2001 and consequent events when the existence of terrorist organizations became obvious to the public. Temporary storage facilities for radioactive wastes may be used as a possible target for terrorist attacks. Untreated...

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

    SciTech Connect (OSTI)

    MANN, F.M.

    2000-03-02

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

  1. Radioactive waste tank Initial Pretreatment Module (IPM) technology development and selection

    SciTech Connect (OSTI)

    Beeman, G.H. [Pacific Northwest Lab., Richland, WA (United States); Hansrote, G. [Westinghouse Hanford Co., Richland, WA (United States)

    1994-03-01

    The processing of nuclear materials at the Hanford Site has resulted in the accumulation of radioactive wastes stored in 177 single- and double-shell tanks (SSTs and DSTs). Fifty-four of the 177 tanks are currently on a tank watch list because organic chemicals and ferrocyanide compounds in the tanks present a potential fire or explosion hazard. In addition, one additional SST is under consideration for placement on the watch list because of high organic concentration. Seventeen of the watch list tanks require pretreatment, and two DST complexant concentrate waste tanks not on the watch list may also need pretreatment. The proposed Initial Pretreatment Module (IPM) is expected to resolve the safety concerns by destroying the organics and ferrocyanide compounds in the tank wastes. The primary objective of the IPM is to destroy or modify constituents that cause safety concerns in the watch list tanks. A secondary objective is to enhance the cost effectiveness of processing the wastes by performing additional processing. Overall, IPM will achieve organic/ferrocyanide destruction (the primary goal) and will assist in the separation of cesium, strontium, and technetium from the tank wastes.

  2. Evaluation of zirconium-iron-rhenium alloys as surrogates for a technetium alloy waste form 

    E-Print Network [OSTI]

    Mews, Paul Aaron

    2008-10-10

    -ray spectroscopy (EDS) capability was employed to determine the phase structure and phase composition of each sample. Iron rich samples were found to form up to three phases, with the rhenium content favoring the intermetallic phases: 1) an Fe solid solution phase...

  3. 324 Building Compliance Project: Selection and evaluation of alternatives for the removal of solid remote-handled mixed wastes from the 324 Building

    SciTech Connect (OSTI)

    Ross, W.A.; Bierschbach, M.C.; Dukelow, J.S. Jr.

    1995-06-01

    Six alternatives for the interim storage of remote-handled mixed wastes from the 324 Building on the Hanford Site have been identified and evaluated. The alternatives focus on the interim storage facility and include use of existing facilities in the 200 Area, the construction of new facilities, and the vitrification of the wastes within the 324 Building to remove the majority of the wastes from under RCRA regulations. The six alternatives are summarized in Table S.1, which identifies the primary facilities to be utilized, the anticipated schedule for removal of the wastes, the costs of the transfer from 324 Building to the interim storage facility (including any capital costs), and an initial risk comparison of the alternatives. A recently negotiated Tri-Party Agreement (TPA) change requires the last of the mixed wastes to be removed by May 1999. The ability to use an existing facility reduces the costs since it eliminates the need for new capital construction. The basic regulatory approvals for the storage of mixed wastes are in place for the PUREX facility, but the Form HI permit will need some minor modifications since the 324 Building wastes have some additional characteristic waste codes and the current permit limits storage of wastes to those from the facility itself. Regulatory reviews have indicated that it will be best to use the tunnels to store the wastes. The PUREX alternatives will only provide storage for about 65% of the wastes. This results from the current schedule of the B-Cell Clean Out Project, which projects that dispersible debris will continue to be collected in small quantities until the year 2000. The remaining fraction of the wastes will then be stored in another facility. Central Waste Complex (CWC) is currently proposed for that residual waste storage; however, other options may also be available.

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

    SciTech Connect (OSTI)

    Mohd Fadzil, Syazwani Binti; Hrma, Pavel R.; Schweiger, Michael J.; Riley, Brian J.

    2015-06-30

    Pyroprocessing is a reprocessing method for managing and reusing used nuclear fuel (UNF) by dissolving it in an electrorefiner with a molten alkali or alkaline earth chloride salt mixture while avoiding wet reprocessing. Pyroprocessing UNF with a LiCl-KCl eutectic salt releases the fission products from the fuel and generates a variety of metallic and salt-based species, including rare earth (RE) chlorides. If the RE-chlorides are converted to oxides, borosilicate glass is a prime candidate for their immobilization because of its durability and ability to dissolve almost any RE waste component into the matrix at high loadings. Crystallization that occurs in waste glasses as the waste loading increases may complicate glass processing and affect the product quality. This work compares three types of borosilicate glasses in terms of liquidus temperature (TL): the International Simple Glass designed by the International Working Group, sodium borosilicate glass developed by Korea Hydro and Nuclear Power, and the lanthanide aluminoborosilicate (LABS) glass established in the United States. The LABS glass allows the highest waste loadings (over 50 mass% RE2O3) while possessing an acceptable chemical durability.

  5. Forms

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would likeUniverse (Journal Article) |FinalIndustrialFollowingForForklift4)Formation offorForms

  6. Forms

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would likeUniverse (Journal Article) |FinalIndustrialFollowingForForklift4)Formation offorForms

  7. Baseline Glass Development for Combined Fission Products Waste Streams

    SciTech Connect (OSTI)

    Crum, Jarrod V.; Billings, Amanda Y.; Lang, Jesse B.; Marra, James C.; Rodriguez, Carmen P.; Ryan, Joseph V.; Vienna, John D.

    2009-06-29

    Borosilicate glass was selected as the baseline technology for immobilization of the Cs/Sr/Ba/Rb (Cs), lanthanide (Ln) and transition metal fission product (TM) waste steams as part of a cost benefit analysis study.[1] Vitrification of the combined waste streams have several advantages, minimization of the number of waste forms, a proven technology, and similarity to waste forms currently accepted for repository disposal. A joint study was undertaken by Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) to develop acceptable glasses for the combined Cs + Ln + TM waste streams (Option 1) and Cs + Ln combined waste streams (Option 2) generated by the AFCI UREX+ set of processes. This study is aimed to develop baseline glasses for both combined waste stream options and identify key waste components and their impact on waste loading. The elemental compositions of the four-corners study were used along with the available separations data to determine the effect of burnup, decay, and separations variability on estimated waste stream compositions.[2-5] Two different components/scenarios were identified that could limit waste loading of the combined Cs + LN + TM waste streams, where as the combined Cs + LN waste stream has no single component that is perceived to limit waste loading. Combined Cs + LN waste stream in a glass waste form will most likely be limited by heat due to the high activity of Cs and Sr isotopes.

  8. GLASS FABRICATION AND ANALYSIS LITERATURE REVIEW AND METHOD SELECTION FOR WTP WASTE FEED QUALIFICATION

    SciTech Connect (OSTI)

    Peeler, D.

    2013-06-27

    Scope of the Report The objective of this literature review is to identify and review documents to address scaling, design, operations, and experimental setup, including configuration, data collection, and remote handling that would be used during waste feed qualification in support of the glass fabrication unit operation. Items addressed include: ? LAW and HLW glass formulation algorithms; ? Mixing and sampling; ? Rheological measurements; ? Heat of hydration; ? Glass fabrication techniques; ? Glass inspection; ? Composition analysis; ? Use of cooling curves; ? Hydrogen generation rate measurement.

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

    SciTech Connect (OSTI)

    Ebert, W. E.

    2006-01-31

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

  10. Selective, nickel-catalyzed carbon-carbon bond-forming reactions of alkynes

    E-Print Network [OSTI]

    Miller, Karen M. (Karen Marie)

    2005-01-01

    Catalytic addition reactions to alkynes are among the most useful and efficient methods for preparing diverse types of substituted olefins. Controlling both regioselectivity and (EIZ)- selectivity in such transformations ...

  11. Remaining Sites Verification Package for the 128-B-2, 100-B Burn Pit #2 Waste Site, Waste Site Reclassification Form 2005-038

    SciTech Connect (OSTI)

    R. A. Carlson

    2005-12-21

    The 128-B-2 waste site was a burn pit historically used for the disposal of combustible and noncombustible wastes, including paint and solvents, office waste, concrete debris, and metallic debris. This site has been remediated by removing approximately 5,627 bank cubic meters of debris, ash, and contaminated soil to the Environmental Restoration Disposal Facility. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  12. DOE Selects Two Small Businesses to Truck Transuranic Waste to New Mexico

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirley Ann JacksonDepartment ofOffice of Headquarters AccountingDOEEnergy DOEDepartmentWaste

  13. A select bibliography with abstracts of reports related to Waste Isolation Pilot Plant geotechnical studies (1972--1990)

    SciTech Connect (OSTI)

    Powers, D.W.; Martin, M.L.

    1993-08-01

    This select bibliography contains 941 entries. Each bibliographic entry contains the citation of a report, conference paper, or journal article containing geotechnical information about the Waste Isolation Pilot Plant (WIPP). The entries cover the period from 1972, when investigation began for a WIPP Site in southeastern New Mexico, through December 1990. Each entry is followed by an abstract. If an abstract or suitable summary existed, it has been included; 316 abstracts were written for other documents. For some entries, an annotation has been provided to clarify the abstract, comment on the setting and significance of the document, or guide the reader to related reports. An index of key words/phrases is included for all entries.

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

    SciTech Connect (OSTI)

    Saueressig, D.G.

    1998-05-20

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

  15. SEISMIC DESIGN REQUIREMENTS SELECTION METHODOLOGY FOR THE SLUDGE TREATMENT & M-91 SOLID WASTE PROCESSING FACILITIES PROJECTS

    SciTech Connect (OSTI)

    RYAN GW

    2008-04-25

    In complying with direction from the U.S. Department of Energy (DOE), Richland Operations Office (RL) (07-KBC-0055, 'Direction Associated with Implementation of DOE-STD-1189 for the Sludge Treatment Project,' and 08-SED-0063, 'RL Action on the Safety Design Strategy (SDS) for Obtaining Additional Solid Waste Processing Capabilities (M-91 Project) and Use of Draft DOE-STD-I 189-YR'), it has been determined that the seismic design requirements currently in the Project Hanford Management Contract (PHMC) will be modified by DOE-STD-1189, Integration of Safety into the Design Process (March 2007 draft), for these two key PHMC projects. Seismic design requirements for other PHMC facilities and projects will remain unchanged. Considering the current early Critical Decision (CD) phases of both the Sludge Treatment Project (STP) and the Solid Waste Processing Facilities (M-91) Project and a strong intent to avoid potentially costly re-work of both engineering and nuclear safety analyses, this document describes how Fluor Hanford, Inc. (FH) will maintain compliance with the PHMC by considering both the current seismic standards referenced by DOE 0 420.1 B, Facility Safety, and draft DOE-STD-1189 (i.e., ASCE/SEI 43-05, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, and ANSI!ANS 2.26-2004, Categorization of Nuclear Facility Structures, Systems and Components for Seismic Design, as modified by draft DOE-STD-1189) to choose the criteria that will result in the most conservative seismic design categorization and engineering design. Following the process described in this document will result in a conservative seismic design categorization and design products. This approach is expected to resolve discrepancies between the existing and new requirements and reduce the risk that project designs and analyses will require revision when the draft DOE-STD-1189 is finalized.

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

    SciTech Connect (OSTI)

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

    2013-12-01

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

  17. Review of Potential Candidate Stabilization Technologies for Liquid and Solid Secondary Waste Streams

    SciTech Connect (OSTI)

    Pierce, Eric M.; Mattigod, Shas V.; Westsik, Joseph H.; Serne, R. Jeffrey; Icenhower, Jonathan P.; Scheele, Randall D.; Um, Wooyong; Qafoku, Nikolla

    2010-01-30

    Pacific Northwest National Laboratory has initiated a waste form testing program to support the long-term durability evaluation of a waste form for secondary wastes generated from the treatment and immobilization of Hanford radioactive tank wastes. The purpose of the work discussed in this report is to identify candidate stabilization technologies and getters that have the potential to successfully treat the secondary waste stream liquid effluent, mainly from off-gas scrubbers and spent solids, produced by the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Down-selection to the most promising stabilization processes/waste forms is needed to support the design of a solidification treatment unit (STU) to be added to the Effluent Treatment Facility (ETF). To support key decision processes, an initial screening of the secondary liquid waste forms must be completed by February 2010.

  18. Hybrid Micro-Electro-Mechanical Systems for Highly Reliable and Selective Characterization of Tank Waste

    SciTech Connect (OSTI)

    Datskos, Panos G.; Sepaniak, Michael J.

    2004-06-01

    Our multifaceted research program is aimed at the fundamental and practical development of hybrid micro-electro-mechanical-systems (MEMS) that integrates several elements of chemical selectivity and sensor function. We are developing MEMS sensors that combine chemimechanical transduction, and surface enhanced Raman spectroscopy (SERS) and radiation detection. One of our goals is to develop highly effective methods of immobilizing a wide variety of molecular and ionic recognition phases onto micromechanical surfaces. We have introduced fundamentally new modes of adsorbate-induced surface stress through nano-structuring of microcantilever surfaces; the responsivity for has increased by over two-orders of magnitude over previously existing technological approaches. Noble metal nanostructures similar to those that enhance chemi-mechanical transduction exhibit substantial Raman enhancement factors.

  19. HerMES: ALMA Imaging of Herschel-selected Dusty Star-forming Galaxies

    E-Print Network [OSTI]

    Bussmann, R S; Fialkov, A; Scudder, J; Hayward, C C; Cowley, W I; Bock, J; Calanog, J; Chapman, S C; Cooray, A; De Bernardis, F; Farrah, D; Fu, Hai; Gavazzi, R; Hopwood, R; Ivison, R J; Jarvis, M; Lacey, C; Loeb, A; Oliver, S J; Perez-Fournon, I; Rigopoulou, D; Roseboom, I G; Scott, Douglas; Smith, A J; Vieira, J D; Wang, L; Wardlow, J

    2015-01-01

    The Herschel Multi-tiered Extragalactic Survey (HerMES) has identified large numbers of dusty star-forming galaxies (DSFGs) over a wide range in redshift. A detailed understanding of these DSFGs is hampered by the poor spatial resolution of Herschel. We present 870um 0.45" imaging obtained in Cycle 0 with the Atacama Large Millimeter/submillimeter Array (ALMA) of a sample of 29 HerMES DSFGs. The ALMA imaging reveals that these DSFGs comprise a total of 62 sources (down to the 5-sigma limit in our ALMA sample; sigma~0.2 mJy). Optical imaging indicates that 36 of the ALMA sources experience a significant flux boost from gravitational lensing (mu>1.1), but only 6 are strongly lensed and show multiple images. We introduce and make use of uvmcmcfit, a general purpose and publicly available Markov chain Monte Carlo visibility plane analysis tool to analyze the source properties. Combined with our previous work on brighter Herschel sources, the lens models presented here tentatively favor intrinsic number counts for...

  20. Remaining Sites Verification Package for the 128-F-3 PNL Burn Pit, Waste Site Reclassification Form 2006-042

    SciTech Connect (OSTI)

    L. M. Dittmer

    2006-10-20

    The 128-F-3 waste site is a former burn pit associated with the 100-F Area experimental animal farm. The site was overlain by coal ash associated with the 126-F-1 waste site and could not be located during confirmatory site evaluation. Therefore, a housekeeping action was performed to remove the coal ash potentially obscuring residual burn pit features. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  1. Evaluation of research and development for terminal isolation of nuclear wastes

    SciTech Connect (OSTI)

    Burton, B.W.

    1982-08-01

    The National Waste Terminal Storage program is responsible for identifying and constructing a geologic repository for spent reactor fuel, high-level waste, and transuranic waste. Extensive research and development work is in progress in the areas of site selection, waste treatment and waste form development, model development and validation, and long-term repository performance assessment. Many potential technologies are under investigation, but specific technologies cannot be identified until a repository site is selected. It is too early in the program to assess the adequacy of environmental control technologies for deep geologic disposal.

  2. Zirconia Inert Matrix Fuel for Plutonium and Minor Actinides Management in Reactors and as an Ultimate Waste Form

    SciTech Connect (OSTI)

    Degueldre, Claude; Wiesenack, Wolfgang

    2008-07-01

    An yttria stabilised zirconia doped with plutonia and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er){sub y}Pu{sub x}Zr{sub 1-y}O{sub 2-{xi}} where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O{sub 2-{xi}} (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia- IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO{sub 2}. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O{sub 2} fuels. The properties of the spent fuel pellets are presented focusing on the once-through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO{sub 2} in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a 'burn and bury' strategy. (authors)

  3. Closed Fuel Cycle Waste Treatment Strategy

    SciTech Connect (OSTI)

    Vienna, J. D.; Collins, E. D.; Crum, J. V.; Ebert, W. L.; Frank, S. M.; Garn, T. G.; Gombert, D.; Jones, R.; Jubin, R. T.; Maio, V. C.; Marra, J. C.; Matyas, J.; Nenoff, T. M.; Riley, B. J.; Sevigny, G. J.; Soelberg, N. R.; Strachan, D. M.; Thallapally, P. K.; Westsik, J. H.

    2015-02-01

    This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

  4. Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank SX-105, Tank AN-103, And AZ-101/102) By Fluidized Bed Steam Reformation (FBSR)

    SciTech Connect (OSTI)

    Jantzen, C. M.; Crawford, C. L.; Bannochie, C. J.; Burket, P. R.; Cozzi, A. D.; Daniel, W. E.; Hall, H. K.; Miller, D. H.; Missimer, D. M.; Nash, C. A.; Williams, M. F.

    2013-09-18

    Fluidized Bed Steam Reforming (FBSR) is a robust technology for the immobilization of a wide variety of radioactive wastes. Applications have been tested at the pilot scale for the high sodium, sulfate, halide, organic and nitrate wastes at the Hanford site, the Idaho National Laboratory (INL), and the Savannah River Site (SRS). Due to the moderate processing temperatures, halides, sulfates, and technetium are retained in mineral phases of the feldspathoid family (nepheline, sodalite, nosean, carnegieite, etc). The feldspathoid minerals bind the contaminants such as Tc-99 in cage (sodalite, nosean) or ring (nepheline) structures to surrounding aluminosilicate tetrahedra in the feldspathoid structures. The granular FBSR mineral waste form that is produced has a comparable durability to LAW glass based on the short term PCT testing in this study, the INL studies, SPFT and PUF testing from previous studies as given in the columns in Table 1-3 that represent the various durability tests. Monolithing of the granular product was shown to be feasible in a separate study. Macro-encapsulating the granular product provides a decrease in leaching compared to the FBSR granular product when the geopolymer is correctly formulated.

  5. Waste Isolation Pilot Plant Transuranic Waste Baseline inventory report. Volume 3. Revision 1

    SciTech Connect (OSTI)

    NONE

    1995-02-01

    This report consists of information related to the waste forms at the WIPP facility from the waste originators. Data for retrievably stored, projected and total wastes are given.

  6. Remaining Sites Verification Package for 132-D-3, 1608-D Effluent Pumping Station, Waste Site Reclassification Form 2005-033

    SciTech Connect (OSTI)

    R. A. Carlson

    2006-05-09

    Decommissioning and demolition of the 132-D-3 site, 1608-D Effluent Pumping Station was performed in 1986. Decommissioning included removal of equipment, water, and sludge for disposal as radioactive waste. The at- and below-grade structure was demolished to at least 1 m below grade and the resulting rubble buried in situ. The area was backfilled to grade with at least 1 m of clean fill and contoured to the surrounding terrain. Residual concentrations support future land uses that can be represented by a rural-residential scenario and pose no threat to groundwater or the Columbia River based on RESRAD modeling.

  7. Disaster waste management: A review article

    SciTech Connect (OSTI)

    Brown, Charlotte; Milke, Mark; Seville, Erica

    2011-06-15

    Depending on their nature and severity, disasters can create large volumes of debris and waste. The waste can overwhelm existing solid waste management facilities and impact on other emergency response and recovery activities. If poorly managed, the waste can have significant environmental and public health impacts and can affect the overall recovery process. This paper presents a system overview of disaster waste management based on existing literature. The main literature available to date comprises disaster waste management plans or guidelines and isolated case studies. There is ample discussion on technical management options such as temporary storage sites, recycling, disposal, etc.; however, there is little or no guidance on how these various management options are selected post-disaster. The literature does not specifically address the impact or appropriateness of existing legislation, organisational structures and funding mechanisms on disaster waste management programmes, nor does it satisfactorily cover the social impact of disaster waste management programmes. It is envisaged that the discussion presented in this paper, and the literature gaps identified, will form a basis for future comprehensive and cohesive research on disaster waste management. In turn, research will lead to better preparedness and response to disaster waste management problems.

  8. Reducing volatilization of heavy metals in phosphate-pretreated municipal solid waste incineration fly ash by forming pyromorphite-like minerals

    SciTech Connect (OSTI)

    Sun Ying; Zheng Jianchang; Zou Luquan; Liu Qiang; Zhu Ping; Qian Guangren

    2011-02-15

    This research investigated the feasibility of reducing volatilization of heavy metals (lead, zinc and cadmium) in municipal solid waste incineration (MSWI) fly ash by forming pyromorphite-like minerals via phosphate pre-treatment. To evaluate the evaporation characteristics of three heavy metals from phosphate-pretreated MSWI fly ash, volatilization tests have been performed by means of a dedicated apparatus in the 100-1000 deg. C range. The toxicity characteristic leaching procedure (TCLP) test and BCR sequential extraction procedure were applied to assess phosphate stabilization process. The results showed that the volatilization behavior in phosphate-pretreated MSWI fly ash could be reduced effectively. Pyromorphite-like minerals formed in phosphate-pretreated MSWI fly ash were mainly responsible for the volatilization reduction of heavy metals in MSWI fly ash at higher temperature, due to their chemical fixation and thermal stabilization for heavy metals. The stabilization effects were encouraging for the potential reuse of MSWI fly ash.

  9. X-ray properties of UV-selected star forming galaxies at z~1 in the Hubble Deep Field North

    E-Print Network [OSTI]

    Laird, E S; Adelberger, K L; Steidel, C C; Reddy, N A

    2005-01-01

    We present an analysis of the X-ray emission from a large sample of ultraviolet (UV) selected, star forming galaxies with 0.74

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

    SciTech Connect (OSTI)

    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. The PA is needed to satisfy both Washington State IDF Permit and DOE Order requirements. Cast Stone has been selected for solidification of radioactive wastes including WTP aqueous secondary wastes treated at the Effluent Treatment Facility (ETF) at Hanford. A similar waste form called Saltstone is used at the Savannah River Site (SRS) to solidify its LAW tank wastes.

  11. Mitigation of Hydrogen Gas Generation from the Reaction of Uranium Metal with Water in K Basin Sludge and Sludge Waste Forms

    SciTech Connect (OSTI)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-08

    Prior laboratory testing identified sodium nitrate and nitrite to be the most promising agents to minimize hydrogen generation from uranium metal aqueous corrosion in Hanford Site K Basin sludge. Of the two, nitrate was determined to be better because of higher chemical capacity, lower toxicity, more reliable efficacy, and fewer side reactions than nitrite. The present lab tests were run to determine if nitrate’s beneficial effects to lower H2 generation in simulated and genuine sludge continued for simulated sludge mixed with agents to immobilize water to help meet the Waste Isolation Pilot Plant (WIPP) waste acceptance drainable liquid criterion. Tests were run at ~60°C, 80°C, and 95°C using near spherical high-purity uranium metal beads and simulated sludge to emulate uranium-rich KW containerized sludge currently residing in engineered containers KW-210 and KW-220. Immobilization agents tested were Portland cement (PC), a commercial blend of PC with sepiolite clay (Aquaset II H), granulated sepiolite clay (Aquaset II G), and sepiolite clay powder (Aquaset II). In all cases except tests with Aquaset II G, the simulated sludge was mixed intimately with the immobilization agent before testing commenced. For the granulated Aquaset II G clay was added to the top of the settled sludge/solution mixture according to manufacturer application directions. The gas volumes and compositions, uranium metal corrosion mass losses, and nitrite, ammonia, and hydroxide concentrations in the interstitial solutions were measured. Uranium metal corrosion rates were compared with rates forecast from the known uranium metal anoxic water corrosion rate law. The ratios of the forecast to the observed rates were calculated to find the corrosion rate attenuation factors. Hydrogen quantities also were measured and compared with quantities expected based on non-attenuated H2 generation at the full forecast anoxic corrosion rate to arrive at H2 attenuation factors. The uranium metal corrosion rates in water alone and in simulated sludge were near or slightly below the metal-in-water rate while nitrate-free sludge/Aquaset II decreased rates by about a factor of 3. Addition of 1 M nitrate to simulated sludge decreased the corrosion rate by a factor of ~5 while 1 M nitrate in sludge/Aquaset II mixtures decreased the corrosion rate by ~2.5 compared with the nitrate-free analogues. Mixtures of simulated sludge with Aquaset II treated with 1 M nitrate had uranium corrosion rates about a factor of 8 to 10 lower than the water-only rate law. Nitrate was found to provide substantial hydrogen mitigation for immobilized simulant sludge waste forms containing Aquaset II or Aquaset II G clay. Hydrogen attenuation factors of 1000 or greater were determined at 60°C for sludge-clay mixtures at 1 M nitrate. Hydrogen mitigation for tests with PC and Aquaset II H (which contains PC) were inconclusive because of suspected failure to overcome induction times and fully enter into anoxic corrosion. Lessening of hydrogen attenuation at ~80°C and ~95°C for simulated sludge and Aquaset II was observed with attenuation factors around 100 to 200 at 1 M nitrate. Valuable additional information has been obtained on the ability of nitrate to attenuate hydrogen gas generation from solution, simulant K Basin sludge, and simulant sludge with immobilization agents. Details on characteristics of the associated reactions were also obtained. The present testing confirms prior work which indicates that nitrate is an effective agent to attenuate hydrogen from uranium metal corrosion in water and simulated K Basin sludge to show that it is also effective in potential candidate solidified K Basin waste forms for WIPP disposal. The hydrogen mitigation afforded by nitrate appears to be sufficient to meet the hydrogen generation limits for shipping various sludge waste streams based on uranium metal concentrations and assumed waste form loadings.

  12. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests

    SciTech Connect (OSTI)

    Thien, Mike G. [Washington River Protection Solutions, LLC, Richland, WA (United States); Barnes, Steve M. [URS, Richland, WA (United States)

    2013-01-17

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described.

  13. Remaining Sites Verification Package for the 100-F-54 Animal Farm Pastures, Waste Site Reclassification Form 2008-015

    SciTech Connect (OSTI)

    J. M. Capron

    2008-04-17

    The 100-F-54 waste site, part of the 100-FR-2 Operable Unit, is the soil associated with the former pastures for holding domestic farm animals used in experimental toxicology studies. Evaluation of historical information resulted in identification of the experimental animal farm pastures as having potential residual soil contamination due to excrement from experimental animals. The 100-F-54 animal farm pastures confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  14. Remaining Sites Verification Package for the 126-F-2, 183-F Clearwells, Waste Site Reclassification Form 2006-017

    SciTech Connect (OSTI)

    R. A. Carlson

    2006-05-04

    The 126-F-2 site is the clearwell facility formerly used as part of the reactor cooling water treatment at the 183-F facility. During demolition operations in the 1970s, potentially contaminated debris was disposed in the eastern clearwell structure. The site has been remediated by removing all debris in the clearwell structure to the Environmental Restoration Disposal Facility. The results of radiological surveys and visual inspection of the remediated clearwell structure show neither residual contamination nor the potential for contaminant migration beyond the clearwell boundaries. The results of verification sampling at the remediation waste staging area demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

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

    SciTech Connect (OSTI)

    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 materials are used. The test cases used in this demonstration are taken from Reference Cases for Use in the Cementitious Barriers Partnership (Langton et al. 2009). Before it is possible to model the release of substances from stabilized waste or radioactive grout through a cement barrier into the engineered soil barrier or natural soil, the relevant characteristics of such materials must be known. Additional chemical characteristics are needed for mechanistic modeling to be undertaken, not just the physical properties relevant for modeling of transport. The minimum required properties for modeling are given in Section 5.0, 'Modeling the chemical speciation of a material'.

  16. Walk the Line: The Development of Route Selection Standards for Spent Nuclear Fuel and High-level Radioactive Waste in the United States - 13519

    SciTech Connect (OSTI)

    Dilger, Fred; Halstead, Robert J.; Ballard, James D.

    2013-07-01

    Although storage facilities for spent nuclear fuel (SNF) and high-level radioactive waste (HLRW) are widely dispersed throughout the United States, these materials are also relatively concentrated in terms of geographic area. That is, the impacts of storage occur in a very small geographic space. Once shipments begin to a national repository or centralized interim storage facility, the impacts of SNF and HLRW will become more geographically distributed, more publicly visible, and almost certainly more contentious. The selection of shipping routes will likely be a major source of controversy. This paper describes the development of procedures, regulations, and standards for the selection of routes used to ship spent nuclear fuel and high-level radioactive waste in the United States. The paper begins by reviewing the circumstances around the development of HM-164 routing guidelines. The paper discusses the significance of New York City versus the Department of Transportation and application of HM-164. The paper describes the methods used to implement those regulations. The paper will also describe the current HM-164 designated routes and will provide a summary data analysis of their characteristics. This analysis will reveal the relatively small spatial scale of the effects of HM 164. The paper will then describe subsequent developments that have affected route selection for these materials. These developments include the use of 'representative routes' found in the Department of Energy (DOE) 2008 Supplemental Environmental Impact Statement for the formerly proposed Yucca Mountain geologic repository. The paper will describe recommendations related to route selection found in the National Academy of Sciences 2006 report Going the Distance, as well as recommendations found in the 2012 Final Report of the Blue Ribbon Commission on America's Nuclear Future. The paper will examine recently promulgated federal regulations (HM-232) for selection of rail routes for hazardous materials transport. The paper concludes that while the HM 164 regime is sufficient for certain applications, it does not provide an adequate basis for a national plan to ship spent nuclear fuel and high-level radioactive waste to centralized storage and disposal facilities over a period of 30 to 50 years. (authors)

  17. CARD No. 24 Waste Characterization

    E-Print Network [OSTI]

    CARD No. 24 Waste Characterization 24.A.1 BACKGROUND DOE must provide waste inventory information Report (TWBIR), Revisions 2 and 3, which provides waste characterization information specific to DOE solidified waste forms was included. Waste described in TWBIR Revision 3 was primarily characterized through

  18. Forms of Soil Phosphorus in Selected Hydrologic Units of the Florida Everglades K. R. Reddy,* Y. Wang, W. F. DeBusk, M. M. Fisher, and S. Newman

    E-Print Network [OSTI]

    Florida, University of

    Forms of Soil Phosphorus in Selected Hydrologic Units of the Florida Everglades K. R. Reddy,* Y is subject to changes in hydroperiod and nutrient loading, resulting in soil P enrichment and changes P in soils from four hydrologic units of the Everglades, and (ii) develop empirical relationships

  19. Comparison of selected DOE and non-DOE requirements, standards, and practices for Low-Level Radioactive Waste Disposal

    SciTech Connect (OSTI)

    Cole, L.; Kudera, D.; Newberry, W.

    1995-12-01

    This document results from the Secretary of Energy`s response to Defense Nuclear Facilities Safety Board Recommendation 94--2. The Secretary stated that the US Department of Energy (DOE) would ``address such issues as...the need for additional requirements, standards, and guidance on low-level radioactive waste management. `` The authors gathered information and compared DOE requirements and standards for the safety aspects Of low-level disposal with similar requirements and standards of non-DOE entities.

  20. Propane reacts with O2 and H2 on gold supported TS-1 to form oxygenates with high selectivity

    E-Print Network [OSTI]

    Bravo Suarez, Juan J.

    2008-06-13

    Gold nanoparticles supported on a microporous titanosilicate (TS-1) were found to be highly selective (95%) towards the formation of acetone and isopropanol from propane, O2, and H2 at moderate temperatures (443 K)....

  1. Cost Comparison for the Transfer of Select Calcined Waste Canisters to the Monitored Geologic Repository at Yucca Mountain, NV

    SciTech Connect (OSTI)

    Michael B. Heiser; Clark B. Millet

    2005-10-01

    This report performs a life-cycle cost comparison of three proposed canister designs for the shipment and disposition of Idaho National Laboratory high-level calcined waste currently in storage at the Idaho Nuclear Technology and Engineering Center to the proposed national monitored geologic repository at Yucca Mountain, Nevada. Concept A (2 × 10-ft) and Concept B (2 × 15-ft) canisters are comparable in design, but they differ in size and waste loading options and vary proportionally in weight. The Concept C (5.5 × 17.5-ft) canister (also called the “super canister”), while similar in design to the other canisters, is considerably larger and heavier than Concept A and B canisters and has a greater wall thickness. This report includes estimating the unique life-cycle costs for the three canister designs. Unique life-cycle costs include elements such as canister purchase and filling at the Idaho Nuclear Technology and Engineering Center, cask preparation and roundtrip consignment costs, final disposition in the monitored geologic repository (including canister off-loading and placement in the final waste disposal package for disposition), and cask purchase. Packaging of the calcine "as-is" would save $2.9 to $3.9 billion over direct vitrification disposal in the proposed national monitored geologic repository at Yucca Mountain, Nevada. Using the larger Concept C canisters would use 0.75 mi less of tunnel space, cost $1.3 billion less than 10-ft canisters of Concept A, and would be complete in 6.2 years.

  2. Nuclear waste package fabricated from concrete

    SciTech Connect (OSTI)

    Pfeiffer, P.A.; Kennedy, J.M.

    1987-03-01

    After the United States enacted the Nuclear Waste Policy Act in 1983, the Department of Energy must design, site, build and operate permanent geologic repositories for high-level nuclear waste. The Department of Energy has recently selected three sites, one being the Hanford Site in the state of Washington. At this particular site, the repository will be located in basalt at a depth of approximately 3000 feet deep. The main concern of this site, is contamination of the groundwater by release of radionuclides from the waste package. The waste package basically has three components: the containment barrier (metal or concrete container, in this study concrete will be considered), the waste form, and other materials (such as packing material, emplacement hole liners, etc.). The containment barriers are the primary waste container structural materials and are intended to provide containment of the nuclear waste up to a thousand years after emplacement. After the containment barriers are breached by groundwater, the packing material (expanding sodium bentonite clay) is expected to provide the primary control of release of radionuclide into the immediate repository environment. The loading conditions on the concrete container (from emplacement to approximately 1000 years), will be twofold; (1) internal heat of the high-level waste which could be up to 400/sup 0/C; (2) external hydrostatic pressure up to 1300 psi after the seepage of groundwater has occurred in the emplacement tunnel. A suggested container is a hollow plain concrete cylinder with both ends capped. 7 refs.

  3. Potential dispositioning flowsheets for ICPP SNF and wastes

    SciTech Connect (OSTI)

    Olson, A.L. [ed.; Anderson, P.A.; Bendixsen, C.L. [and others

    1995-11-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation`s radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995.

  4. Insurance Enrollment Form -2014 Plan Year -New Enrollees Select one: Norman Campus Oklahoma City Campus Tulsa Campus

    E-Print Network [OSTI]

    Oklahoma, University of

    Campus Tulsa Campus Complete all sections of this form and return this original signed document to your - (405) 271-2180. Tulsa Campus - Office of Human Resources - 4502 E. 41st. Street, Suite 1C114 - Tulsa

  5. Using geologic conditions and multiattribute decision analysis to determine the relative favorability of selected areas for siting a high-level radioactive waste repository

    SciTech Connect (OSTI)

    Harrison, W.; Edgar, D.E.; Baker, C.H.; Buehring, W.A.; Whitfield, R.G.; Van Luik, A.E.J.; Sood, M.K.; Flower, M.F.J.; Warren, M.F.; Jusko, M.J.; Peerenboom, J.P.; Bogner, J.E.

    1988-05-01

    A method is presented for determining the relative favorability of geologically complex areas for isolating high-level radioactive wastes. In applying the method to the northeastern region of the United States, seismicity and tectonic activity were the screening criteria used to divide the region into three areas of increasing seismotectonic risk. Criteria were then used to subdivide the area of lowest seismotectonic risk into six geologically distinct subareas including characteristics, surface-water and groundwater hydrology, potential human intrusion, site geometry, surface characteristics, and tectonic environment. Decision analysis was then used to identify the subareas most favorable from a geologic standpoint for further investigation, with a view to selecting a site for a repository. Three subareas (parts of northeastern Vermont, northern New Hampshire, and western Maine) were found to be the most favorable, using this method and existing data. However, because this study assessed relative geologic favorability, no conclusions should be drawn concerning the absolute suitability of individual subareas for high-level radioactive waste isolation. 34 refs., 7 figs., 20 tabs.

  6. Underground waste barrier structure

    DOE Patents [OSTI]

    Saha, Anuj J. (Hamburg, NY); Grant, David C. (Gibsonia, PA)

    1988-01-01

    Disclosed is an underground waste barrier structure that consists of waste material, a first container formed of activated carbonaceous material enclosing the waste material, a second container formed of zeolite enclosing the first container, and clay covering the second container. The underground waste barrier structure is constructed by forming a recessed area within the earth, lining the recessed area with a layer of clay, lining the clay with a layer of zeolite, lining the zeolite with a layer of activated carbonaceous material, placing the waste material within the lined recessed area, forming a ceiling over the waste material of a layer of activated carbonaceous material, a layer of zeolite, and a layer of clay, the layers in the ceiling cojoining with the respective layers forming the walls of the structure, and finally, covering the ceiling with earth.

  7. Waste disposal package

    DOE Patents [OSTI]

    Smith, M.J.

    1985-06-19

    This is a claim for a waste disposal package including an inner or primary canister for containing hazardous and/or radioactive wastes. The primary canister is encapsulated by an outer or secondary barrier formed of a porous ceramic material to control ingress of water to the canister and the release rate of wastes upon breach on the canister. 4 figs.

  8. Selective Capture of Cesium and Thallium from Natural Waters and Simulated Wastes with Copper Ferrocyanide Functionalized Mesoporous Silica

    SciTech Connect (OSTI)

    Sangvanich, Thanapon; Sukwarotwat, Vichaya; Wiacek, Robert J.; Grudzien, Rafal M.; Fryxell, Glen E.; Addleman, Raymond S.; Timchalk, Charles; Yantasee, Wassana

    2010-10-15

    Copper(II) ferrocyanide immobilized inside mesoporous silica MCM-41 supports (Cu-FC-EDA-SAMMSTM) has been evaluated against iron(III) hexacyanoferrate(II) (insoluble Prussian blue) for the sorption of cesium (Cs+) and thallium (Tl+) from natural waters and simulated wastes. The affinities (in term of distribution coefficients, Kd) of both sorbents for Cs and Tl were measured as a function of solution pH, competing cations, and matrices. For the entire pH studied (pH 0.1 to 7.3), Cu-FC-EDA-SAMMS had higher affinities for Cs and Tl (one to two orders of magnitude higher Kd) than Prussian blue and was less negatively impacted by the solution pH, competing cations, and matrices. The adsorption isotherms and kinetics of the two sorbents for Cs and/or Tl were also determined in seawater and simulated acid and alkaline wastes. SAMMS outperformed Prussian blue in terms of maximum adsorption capacity (e.g., 21.7 versus 2.6 mg Cs/g in acid waste stimulant, pH 1.1), and rate (e.g., over 95 wt% of Cs was removed after 2 minutes with SAMMS, while only 75 wt% was removed with Prussian blue). The lower affinity, capacity, and rate of Cs and Tl sorption on Prussian blue than those on Cu-FC-EDA-SAMMS were attributed to the molecular pore sizes, which restrict mass transport, and the insoluble Cs abducts of the Prussian blue, which restrict the ability of neighboring binding sites to further bind Cs ions. On the other hand, the large pores of SAMMS not only enable faster diffusion and faster binding chemistry, but they also allow isolation of binding sites so that one Cs binding event does not impact further Cs binding. In addition, iron (Fe) dissolved from insoluble Prussian blue over 10-fold of that from Cu-FC-EDA-SAMMS after 24 hours of contact time, indicating poorer material stability of Prussian blue.

  9. Preliminary assessment of candidate immobilization technologies for retrieved single-shell tank wastes

    SciTech Connect (OSTI)

    Wiemers, K.D.; Mendel, J.E.; Kruger, A.A.; Bunnell, L.R.; Mellinger, G.B.

    1992-01-01

    This report describes the initial work that has been performed to select technologies for immobilization of wastes that may be retrieved from Hanford single-shell tanks (SSTs). Two classes of waste will require immobilization. One is the combined high-level waste/transuranic (HLW/TRU) fraction, the other the low-level waste (LLW) fraction. A number of potential immobilization technologies are identified for each class of waste. Immobilization technologies were initially selected based on a number of considerations, including (1) the waste loading that could likely be achieved within the constraint of producing acceptable waste forms, (2) process flexibility (primarily compatibility with anticipated waste variability), (3) process complexity, and (4) state of development. Immobilization technologies selected for further development include the following: for HLW/TRU waste -- borosilicate glass, lead-iron phosphate glass, glass-calcine composites, glass-ceramics, and cement based forms; for non-denitrated LLW -- grout, laxtex-modified concrete, and polyethylene; and for denitrated LLW -- silicate glass, phosphate glass, and clay calcination or tailored ceramic in various matrices.

  10. Kaiser Engineers Hanford internal position paper -- Project W-236A, Multi-function Waste Tank Facility -- Peer reviews of selected activities

    SciTech Connect (OSTI)

    Stine, M.D.

    1995-01-04

    The purpose of this paper is to develop and document a proposed position on the performance of independent peer reviews on selected design and analysis components of the Title 1 [Preliminary] and Title 2 [Final] design phases of the Multi-Function Waste Tank Facility [MWTF] project. An independent, third-party peer review is defined as a documented critical review of documents, data, designs, design inputs, tests, calculations, or related materials. The peer review should be conducted by persons independent of those who performed the work, but who are technically qualified to perform the original work. The peer review is used to assess the validity of assumptions and functional requirements, to assess the appropriateness and logic of selected methodologies and design inputs, and to verify calculations, analyses and computer software. The peer review can be conducted at the end of the design activity, at specific stages of the design process, or continuously and concurrently with the design activity. This latter method is often referred to as ``Continuous Peer Review.``

  11. Advanced radioactive waste-glass melters

    SciTech Connect (OSTI)

    Bickford, D.F.

    1990-12-31

    During pilot scale operations of the Scale Glass Melter for the US Department of Energy a team of engineers and scientists was formed to assess the need for continued melter design development to support the Defense Waste Processing Facility (DWPF), and prioritize future efforts. Recently this has taken on new importance because of selection of the DWPF Melter design as the reference for the Hanford Waste Vitrification Project (HWVP), and increased interest at the West Valley Demonstration Project on melter life and replacement. Results of the study are summarized, and goals produced by the study are compared to the results of current programs at the Savannah River Laboratory (SRL).

  12. Advanced radioactive waste-glass melters

    SciTech Connect (OSTI)

    Bickford, D.F.

    1990-01-01

    During pilot scale operations of the Scale Glass Melter for the US Department of Energy a team of engineers and scientists was formed to assess the need for continued melter design development to support the Defense Waste Processing Facility (DWPF), and prioritize future efforts. Recently this has taken on new importance because of selection of the DWPF Melter design as the reference for the Hanford Waste Vitrification Project (HWVP), and increased interest at the West Valley Demonstration Project on melter life and replacement. Results of the study are summarized, and goals produced by the study are compared to the results of current programs at the Savannah River Laboratory (SRL).

  13. West Valley demonstration project: alternative processes for solidifying the high-level wastes

    SciTech Connect (OSTI)

    Holton, L.K.; Larson, D.E.; Partain, W.L.; Treat, R.L.

    1981-10-01

    In 1980, the US Department of Energy (DOE) established the West Valley Solidification Project as the result of legislation passed by the US Congress. The purpose of this project was to carry out a high level nuclear waste management demonstration project at the Western New York Nuclear Service Center in West Valley, New York. The DOE authorized the Pacific Northwest Laboratory (PNL), which is operated by Battelle Memorial Institute, to assess alternative processes for treatment and solidification of the WNYNSC high-level wastes. The Process Alternatives Study is the suject of this report. Two pretreatment approaches and several waste form processes were selected for evaluation in this study. The two waste treatment approaches were the salt/sludge separation process and the combined waste process. Both terminal and interim waste form processes were studied.

  14. SELECTIVE REMOVAL OF STRONTIUM AND CESIUM FROM SIMULATED WASTE SOLUTION WITH TITANATE ION-EXCHANGERS IN A FILTER CARTRIDGE CONFIGURATIONS-12092

    SciTech Connect (OSTI)

    Oji, L.; Martin, K.; Hobbs, D.

    2012-01-03

    Experimental results for the selective removal of strontium and cesium from simulated waste solutions with monosodium titanate and crystalline silicotitanate laden filter cartridges are presented. In these proof-of-principle tests, effective uptake of both strontium-85 and cesium-137 were observed using ion-exchangers in this filter cartridge configuration. At low salt simulant conditions, the instantaneous decontamination factor for strontium-85 with monosodium titanate impregnated filter membrane cartridges measured 26, representing 96% strontium-85 removal efficiency. On the other hand, the strontium-85 instantaneous decontamination factor with co-sintered active monosodium titanate cartridges measured 40 or 98% Sr-85 removal efficiency. Strontium-85 removal with the monosodium titanate impregnated membrane cartridges and crystalline silicotitanate impregnated membrane cartridges, placed in series arrangement, produced an instantaneous decontamination factor of 41 compared to an instantaneous decontamination factor of 368 for strontium-85 with co-sintered active monosodium titanate cartridges and co-sintered active crystalline silicotitanate cartridges placed in series. Overall, polyethylene co-sintered active titanates cartridges performed as well as titanate impregnated filter membrane cartridges in the uptake of strontium. At low ionic strength conditions, there was a significant uptake of cesium-137 with co-sintered crystalline silicotitanate cartridges. Tests results with crystalline silicotitanate impregnated membrane cartridges for cesium-137 decontamination are currently being re-evaluated. Based on these preliminary findings we conclude that incorporating monosodium titanate and crystalline silicotitanate sorbents into membranes represent a promising method for the semicontinuous removal of radioisotopes of strontium and cesium from nuclear waste solutions.

  15. SELECTIVE REMOVAL OF STRONTIUM AND CESIUM FROM SIMULATED WASTE SOLUTION WITH TITANATE ION-EXCHANGERS IN A FILTER CARTRIDGE CONFIGURATIONS-12092

    SciTech Connect (OSTI)

    Oji, L.; Martin, K.; Hobbs, D.

    2011-11-10

    Experimental results for the selective removal of strontium and cesium from simulated waste solutions with monosodium titanate (MST) and crystalline silicotitanate (CST) laden filter cartridges are presented. In these proof-of-principle tests, effective uptake of both Sr-85 and Cs-137 were observed using ion-exchangers in this filter cartridge configuration. At low salt simulant conditions, the instantaneous decontamination factor (D{sub F}) for Sr-85 with MST impregnated filter membrane cartridges measured 26, representing 96% Sr-85 removal efficiency. On the other hand, the Sr-85 instantaneous D{sub F} with co-sintered active MST cartridges measured 40 or 98% Sr-85 removal efficiency. Strontium-85 removal with the MST impregnated membrane cartridges and CST impregnated membrane cartridges, placed in series arrangement, produced an instantaneous decontamination factor of 41 compared to an instantaneous decontamination factor of 368 for strontium-85 with co-sintered active MST cartridges and co-sintered active CST cartridges placed in series. Overall, polyethylene co-sintered active titanates cartridges performed as well as titanate impregnated filter membrane cartridges in the uptake of strontium. At low ionic strength conditions, there was a significant uptake of Cs-137 with co-sintered CST cartridges. Tests results with CST impregnated membrane cartridges for Cs-137 decontamination are currently being re-evaluated. Based on these preliminary findings we conclude that incorporating MST and CST sorbents into membranes represent a promising method for the semi-continuous removal of radioisotopes of strontium and cesium from nuclear waste solutions.

  16. Stabilization of vitrified wastes: Task 4. Topical report, October 1994--September 1995

    SciTech Connect (OSTI)

    Nowok, J.W.; Pflughoeft-Hassett, D.F.; Hassett, D.J.; Hurley, J.P.

    1995-09-01

    The goal of this task was to work with private industry to refine existing vitrification processes to produce a more stable vitrified product. The initial objectives were to (1) demonstrate a waste vitrification procedure for enhanced stabilization of waste materials and (2) develop a testing protocol to understand the long-term leaching behavior of the stabilized waste form. The testing protocol was expected to be based on a leaching procedure called the synthetic groundwater leaching procedure (SGLP). This task will contribute to the US DOE`s identified technical needs in waste characterization, low-level mixed-waste processing, disposition technology, and improved waste forms. The proposed work was to proceed over 4 years in the following steps: literature surveys to aid in the selection and characterization of test mixtures for vitrification, characterization of optimized vitrified test wastes using advanced leaching protocols, and refinement and demonstration of vitrification methods leading to commercialization. For this year, literature surveys were completed, and computer modeling was performed to determine the feasibility of removing heavy metals from a waste during vitrification, thereby reducing the hazardous nature of the vitrified material and possibly producing a commercial metal concentrate. This report describes the following four subtasks: survey of vitrification technologies; survey of cleanup sites; selection and characterization of test mixtures for vitrification and crystallization; and selection of crystallization methods based on thermochemistry modeling.

  17. Hazardous Waste Management Overview The Five L's

    E-Print Network [OSTI]

    Jia, Songtao

    Hazardous Waste Management Overview The Five L's CoLLect CoLLect all hazardous chemical waste and submit a chemical waste pick-up request form for proper disposal. Periodically evaluate your chemical are unsure if your chemical waste is a Hazardous Waste, consult EH&S at hazmat@columbia.edu. DO

  18. Iron phosphate compositions for containment of hazardous metal waste

    DOE Patents [OSTI]

    Day, D.E.

    1998-05-12

    An improved iron phosphate waste form for the vitrification, containment and long-term disposition of hazardous metal waste such as radioactive nuclear waste is provided. The waste form comprises a rigid iron phosphate matrix resulting from the cooling of a melt formed by heating a batch mixture comprising the metal waste and a matrix-forming component. The waste form comprises from about 30 to about 70 weight percent P{sub 2}O{sub 5} and from about 25 to about 50 weight percent iron oxide and has metals present in the metal waste chemically dissolved therein. The concentration of iron oxide in the waste form along with a high proportion of the iron in the waste form being present as Fe{sup 3+} provide a waste form exhibiting improved chemical resistance to corrosive attack. A method for preparing the improved iron phosphate waste forms is also provided. 21 figs.

  19. Iron phosphate compositions for containment of hazardous metal waste

    DOE Patents [OSTI]

    Day, Delbert E. (Rolla, MO)

    1998-01-01

    An improved iron phosphate waste form for the vitrification, containment and long-term disposition of hazardous metal waste such as radioactive nuclear waste is provided. The waste form comprises a rigid iron phosphate matrix resulting from the cooling of a melt formed by heating a batch mixture comprising the metal waste and a matrix-forming component. The waste form comprises from about 30 to about 70 weight percent P.sub.2 O.sub.5 and from about 25 to about 50 weight percent iron oxide and has metals present in the metal waste chemically dissolved therein. The concentration of iron oxide in the waste form along with a high proportion of the iron in the waste form being present as Fe.sup.3+ provide a waste form exhibiting improved chemical resistance to corrosive attack. A method for preparing the improved iron phosphate waste forms is also provided.

  20. SUMMARY PLAN FOR BENCH-SCALE REFORMER AND PRODUCT TESTING TREATABILITY STUDIES USING HANFORD TANK WASTE

    SciTech Connect (OSTI)

    DUNCAN JB

    2010-08-19

    This paper describes the sample selection, sample preparation, environmental, and regulatory considerations for shipment of Hanford radioactive waste samples for treatability studies of the FBSR process at the Savannah River National Laboratory and the Pacific Northwest National Laboratory. The U.S. Department of Energy (DOE) Hanford tank farms contain approximately 57 million gallons of wastes, most of which originated during the reprocessing of spent nuclear fuel to produce plutonium for defense purposes. DOE intends to pre-treat the tank waste to separate the waste into a high level fraction, that will be vitrified and disposed of in a national repository as high-level waste (HLW), and a low-activity waste (LAW) fraction that will be immobilized for on-site disposal at Hanford. The Hanford Waste Treatment and Immobilization Plant (WTP) is the focal point for the treatment of Hanford tank waste. However, the WTP lacks the capacity to process all of the LAW within the regulatory required timeframe. Consequently, a supplemental LAW immobilization process will be required to immobilize the remainder of the LAW. One promising supplemental technology is Fluidized Bed Steam Reforming (FBSR) to produce a sodium-alumino-silicate (NAS) waste form. The NAS waste form is primarily composed of nepheline (NaAlSiO{sub 4}), sodalite (Nas[AlSiO{sub 4}]{sub 6}Cl{sub 2}), and nosean (Na{sub 8}[AlSiO{sub 4}]{sub 6}SO{sub 4}). Semivolatile anions such as pertechnetate (TcO{sub 4}{sup -}) and volatiles such as iodine as iodide (I{sup -}) are expected to be entrapped within the mineral structures, thereby immobilizing them (Janzen 2008). Results from preliminary performance tests using surrogates, suggests that the release of semivolatile radionuclides {sup 99}Tc and volatile {sup 129}I from granular NAS waste form is limited by Nosean solubility. The predicted release of {sup 99}Tc from the NAS waste form at a 100 meters down gradient well from the Integrated Disposal Facility (IDF) was found to be comparable to immobilized low-activity waste glass waste form in the initial supplemental LAW treatment technology risk assessment (Mann 2003). To confirm this hypothesis, DOE is funding a treatability study where three actual Hanford tank waste samples (containing both {sup 99}Tc and {sup 125}I) will be processed in Savannah River National Laboratory's (SRNL) Bench-Scale Reformer (BSR) to form the mineral product, similar to the granular NAS waste form, that will then be subject to a number of waste form qualification tests. In previous tests, SRNL have demonstrated that the BSR product is chemically and physically equivalent to the FBSR product (Janzen 2005). The objective of this paper is to describe the sample selection, sample preparation, and environmental and regulatory considerations for treatability studies of the FBSR process using Hanford tank waste samples at the SNRL. The SNRL will process samples in its BSR. These samples will be decontaminated in the 222-S Laboratory to remove undissolved solids and selected radioisotopes to comply with Department of Transportation (DOT) shipping regulations and to ensure worker safety by limiting radiation exposure to As Low As Reasonably Achievable (ALARA). These decontamination levels will also meet the Nuclear Regulatory Commission's (NRC's) definition of low activity waste (LAW). After the SNRL has processed the tank samples to a granular mineral form, SRNL and Pacific Northwest National Laboratory (PNNL) will conduct waste form testing on both the granular material and monoliths prepared from the granular material. The tests being performed are outlined in Appendix A.

  1. Melt Processed Single Phase Hollandite Waste Forms For Nuclear Waste Immobilization: Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al

    SciTech Connect (OSTI)

    Brinkman, Kyle [Savannah River National Laboratory, Aiken, SC 29808 (United States); Marra, James [Savannah River National Laboratory, Aiken, SC 29808 (United States); Amoroso, Jake [Savannah River National Laboratory, Aiken, SC 29808 (United States); Conradson, Steven D. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Tang, Ming [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-09-23

    Cs is one of the more problematic fission product radionuclides to immobilize due to its high volatility at elevated temperatures, ability to form water soluble compounds, and its mobility in many host materials. The hollandite structure is a promising crystalline host for Cs immobilization and has been traditionally fabricated by solid state sintering methods. This study presents the structure and performance of Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al hollandite fabricated by melt processing. Melt processing is considered advantageous given that melters are currently in use for High Level Waste (HLW) vitrification in several countries. This work details the impact of Cr additions that were demonstrated to i) promote the formation of a Cs containing hollandite phase and ii) maintain the stability of the hollandite phase in reducing conditions anticipated for multiphase waste form processing.

  2. Radioactive waste disposal package

    DOE Patents [OSTI]

    Lampe, Robert F. (Bethel Park, PA)

    1986-01-01

    A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

  3. Apparatus for forming thin-film heterojunction solar cells employing materials selected from the class of I-III-VI.sub.2 chalcopyrite compounds

    DOE Patents [OSTI]

    Mickelsen, Reid A. (Bellevue, WA); Chen, Wen S. (Seattle, WA)

    1983-01-01

    Apparatus for forming thin-film, large area solar cells having a relatively high light-to-electrical energy conversion efficiency and characterized in that the cell comprises a p-n-type heterojunction formed of: (i) a first semiconductor layer comprising a photovoltaic active material selected from the class of I-III-VI.sub.2 chalcopyrite ternary materials which is vacuum deposited in a thin "composition-graded" layer ranging from on the order of about 2.5 microns to about 5.0 microns (.congruent.2.5 .mu.m to .congruent.5.0 .mu.m) and wherein the lower region of the photovoltaic active material preferably comprises a low resistivity region of p-type semiconductor material having a superimposed region of relatively high resistivity, transient n-type semiconductor material defining a transient p-n homojunction; and (ii), a second semiconductor layer comprising a low resistivity n-type semiconductor material wherein interdiffusion (a) between the elemental constituents of the two discrete juxtaposed regions of the first semiconductor layer defining a transient p-n homojunction layer, and (b) between the transient n-type material in the first semiconductor layer and the second n-type semiconductor layer, causes the transient n-type material in the first semiconductor layer to evolve into p-type material, thereby defining a thin layer heterojunction device characterized by the absence of voids, vacancies and nodules which tend to reduce the energy conversion efficiency of the system.

  4. Treatment of mercury containing waste

    DOE Patents [OSTI]

    Kalb, Paul D. (Wading River, NY); Melamed, Dan (Gaithersburg, MD); Patel, Bhavesh R (Elmhurst, NY); Fuhrmann, Mark (Babylon, NY)

    2002-01-01

    A process is provided for the treatment of mercury containing waste in a single reaction vessel which includes a) stabilizing the waste with sulfur polymer cement under an inert atmosphere to form a resulting mixture and b) encapsulating the resulting mixture by heating the mixture to form a molten product and casting the molten product as a monolithic final waste form. Additional sulfur polymer cement can be added in the encapsulation step if needed, and a stabilizing additive can be added in the process to improve the leaching properties of the waste form.

  5. Method for solidification of radioactive and other hazardous waste

    DOE Patents [OSTI]

    Anshits, Alexander G. (Krasnoyarsk, RU); Vereshchagina, Tatiana A. (Krasnoyarsk, RU); Voskresenskaya, Elena N. (Krasnoyarsk, RU); Kostin, Eduard M. (Zheleznogorsk, RU); Pavlov, Vyacheslav F. (Krasnoyarsk, RU); Revenko, Yurii A. (Zheleznogorsk, RU); Tretyakov, Alexander A. (Zheleznogorsk, RU); Sharonova, Olga M. (Krasnoyarsk, RU); Aloy, Albert S. (Saint-Petersburg, RU); Sapozhnikova, Natalia V. (Saint-Petersburg, RU); Knecht, Dieter A. (Idaho Falls, ID); Tranter, Troy J. (Idaho Falls, ID); Macheret, Yevgeny (Idaho Falls, ID)

    2002-01-01

    Solidification of liquid radioactive waste, and other hazardous wastes, is accomplished by the method of the invention by incorporating the waste into a porous glass crystalline molded block. The porous block is first loaded with the liquid waste and then dehydrated and exposed to thermal treatment at 50-1,000.degree. C. The porous glass crystalline molded block consists of glass crystalline hollow microspheres separated from fly ash (cenospheres), resulting from incineration of fossil plant coals. In a preferred embodiment, the porous glass crystalline blocks are formed from perforated cenospheres of grain size -400+50, wherein the selected cenospheres are consolidated into the porous molded block with a binder, such as liquid silicate glass. The porous blocks are then subjected to repeated cycles of saturating with liquid waste, and drying, and after the last cycle the blocks are subjected to calcination to transform the dried salts to more stable oxides. Radioactive liquid waste can be further stabilized in the porous blocks by coating the internal surface of the block with metal oxides prior to adding the liquid waste, and by coating the outside of the block with a low-melting glass or a ceramic after the waste is loaded into the block.

  6. A Case for Molecular Recognition in Nuclear Separations: Sulfate Separation from Nuclear Wastes

    SciTech Connect (OSTI)

    Moyer, Bruce A; Custelcean, Radu; Hay, Benjamin; Sessler, Jonathan L.; Bowman-James, Kristin; Day, Victor W.; Kang, S.O.

    2013-01-01

    In this paper, we present the case for molecular-recognition approaches for sulfate removal from radioactive wastes via the use of anion-sequestering systems selective for sulfate, using either liquid liquid extraction or crystallization. Potential benefits of removing sulfate from the waste include improved vitrification of the waste, reduced waste-form volume, and higher waste-form performance, all of which lead to potential cleanup schedule acceleration and cost savings. The need for sulfate removal from radioactive waste, especially legacy tank wastes stored at the Hanford site, is reviewed in detail and primarily relates to the low solubility of sulfate in borosilicate glass. Traditional methods applicable to the separation of sulfate from radioactive wastes are also reviewed, with the finding that currently no technology has been identified and successfully demonstrated to meet this need. Fundamental research in the authors laboratories targeting sulfate as an important representative of the class of oxoanions is based on the hypothesis that designed receptors may provide the needed ability to recognize sulfate under highly competitive conditions, in particular where the nitrate anion concentration is high. Receptors that have been shown to have promising affinity for sulfate, either in extraction or in crystallization experiments, include hexaurea tripods, tetraamide macrocycles, cyclo[8]pyrroles, calixpyrroles, and self-assembled urea-lined cages. Good sulfate selectivity observed in the laboratory provides experimental support for the proposed molecular-recognition approach.

  7. Vitrification of hazardous and radioactive wastes

    SciTech Connect (OSTI)

    Bickford, D.F.; Schumacher, R.

    1995-12-31

    Vitrification offers many attractive waste stabilization options. Versatility of waste compositions, as well as the inherent durability of a glass waste form, have made vitrification the treatment of choice for high-level radioactive wastes. Adapting the technology to other hazardous and radioactive waste streams will provide an environmentally acceptable solution to many of the waste challenges that face the public today. This document reviews various types and technologies involved in vitrification.

  8. Remaining Sites Verification Package for the 100-F-26:13, 108-F Drain Pipelines, Waste Site Reclassification Form 2005-011

    SciTech Connect (OSTI)

    L. M. Dittmer

    2008-03-03

    The 100-F-26:13 waste site is the network of process sewer pipelines that received effluent from the 108-F Biological Laboratory and discharged it to the 188-F Ash Disposal Area (126-F-1 waste site). The pipelines included one 0.15-m (6-in.)-, two 0.2-m (8-in.)-, and one 0.31-m (12-in.)-diameter vitrified clay pipe segments encased in concrete. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  9. Remaining Sites Verification Package for the 1607-B2 Septic System and 100-B-14:2 Sanitary Sewer System, Waste Site Reclassification Form 2006-055

    SciTech Connect (OSTI)

    L. M. Dittmer

    2007-03-21

    The 1607-B2 waste site is a former septic system associated with various 100-B facilities, including the 105-B, 108-B, 115-B/C, and 185/190-B buildings. The site was evaluated based on confirmatory results for feeder lines within the 100-B-14:2 subsite and determined to require remediation. The 1607-B2 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  10. Low-level tank waste simulant data base

    SciTech Connect (OSTI)

    Lokken, R.O.

    1996-04-01

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

  11. Municipal solid waste combustion: Fuel testing and characterization

    SciTech Connect (OSTI)

    Bushnell, D.J.; Canova, J.H.; Dadkhah-Nikoo, A.

    1990-10-01

    The objective of this study is to screen and characterize potential biomass fuels from waste streams. This will be accomplished by determining the types of pollutants produced while burning selected municipal waste, i.e., commercial mixed waste paper residential (curbside) mixed waste paper, and refuse derived fuel. These materials will be fired alone and in combination with wood, equal parts by weight. The data from these experiments could be utilized to size pollution control equipment required to meet emission standards. This document provides detailed descriptions of the testing methods and evaluation procedures used in the combustion testing and characterization project. The fuel samples will be examined thoroughly from the raw form to the exhaust emissions produced during the combustion test of a densified sample.

  12. WASTE PACKAGE TRANSPORTER DESIGN

    SciTech Connect (OSTI)

    D.C. Weddle; R. Novotny; J. Cron

    1998-09-23

    The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''.

  13. Idaho Waste Vitrification Facilities Project Vitrified Waste Interim Storage Facility

    SciTech Connect (OSTI)

    Bonnema, Bruce Edward

    2001-09-01

    This feasibility study report presents a draft design of the Vitrified Waste Interim Storage Facility (VWISF), which is one of three subprojects of the Idaho Waste Vitrification Facilities (IWVF) project. The primary goal of the IWVF project is to design and construct a treatment process system that will vitrify the sodium-bearing waste (SBW) to a final waste form. The project will consist of three subprojects that include the Waste Collection Tanks Facility, the Waste Vitrification Facility (WVF), and the VWISF. The Waste Collection Tanks Facility will provide for waste collection, feed mixing, and surge storage for SBW and newly generated liquid waste from ongoing operations at the Idaho Nuclear Technology and Engineering Center. The WVF will contain the vitrification process that will mix the waste with glass-forming chemicals or frit and turn the waste into glass. The VWISF will provide a shielded storage facility for the glass until the waste can be disposed at either the Waste Isolation Pilot Plant as mixed transuranic waste or at the future national geological repository as high-level waste glass, pending the outcome of a Waste Incidental to Reprocessing determination, which is currently in progress. A secondary goal is to provide a facility that can be easily modified later to accommodate storage of the vitrified high-level waste calcine. The objective of this study was to determine the feasibility of the VWISF, which would be constructed in compliance with applicable federal, state, and local laws. This project supports the Department of Energy’s Environmental Management missions of safely storing and treating radioactive wastes as well as meeting Federal Facility Compliance commitments made to the State of Idaho.

  14. Remaining Sites Verification Package for the 100-F-26:15 Miscellaneous Pipelines Associated with the 132-F-6, 1608-F Waste Water Pumping Station, Waste Site Reclassification Form 2007-031

    SciTech Connect (OSTI)

    L. M. Dittmer

    2008-03-18

    The 100-F-26:15 waste site consisted of the remnant portions of underground process effluent and floor drain pipelines that originated at the 105-F Reactor. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  15. Mixed waste characterization reference document

    SciTech Connect (OSTI)

    1997-09-01

    Waste characterization and monitoring are major activities in the management of waste from generation through storage and treatment to disposal. Adequate waste characterization is necessary to ensure safe storage, selection of appropriate and effective treatment, and adherence to disposal standards. For some wastes characterization objectives can be difficult and costly to achieve. The purpose of this document is to evaluate costs of characterizing one such waste type, mixed (hazardous and radioactive) waste. For the purpose of this document, waste characterization includes treatment system monitoring, where monitoring is a supplement or substitute for waste characterization. This document establishes a cost baseline for mixed waste characterization and treatment system monitoring requirements from which to evaluate alternatives. The cost baseline established as part of this work includes costs for a thermal treatment technology (i.e., a rotary kiln incinerator), a nonthermal treatment process (i.e., waste sorting, macronencapsulation, and catalytic wet oxidation), and no treatment (i.e., disposal of waste at the Waste Isolation Pilot Plant (WIPP)). The analysis of improvement over the baseline includes assessment of promising areas for technology development in front-end waste characterization, process equipment, off gas controls, and monitoring. Based on this assessment, an ideal characterization and monitoring configuration is described that minimizes costs and optimizes resources required for waste characterization.

  16. Nevada Industrial Solid Waste Disposal Site Permit Application...

    Open Energy Info (EERE)

    Nevada Industrial Solid Waste Disposal Site Permit Application Jump to: navigation, search OpenEI Reference LibraryAdd to library Form: Nevada Industrial Solid Waste Disposal Site...

  17. Performance assessment model of a single waste package (Conference...

    Office of Scientific and Technical Information (OSTI)

    12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; WASTE FORMS; PERFORMANCE; P CODES; RADIOACTIVE...

  18. Preliminary analysis of treatment strategies for transuranic wastes from reprocessing plants

    SciTech Connect (OSTI)

    Ross, W.A.; Schneider, K.J.; Swanson, J.L.; Yasutake, K.M.; Allen, R.P.

    1985-07-01

    This document provides a comparison of six treatment options for transuranic wastes (TRUW) resulting from the reprocessing of commercial spent fuel. Projected transuranic waste streams from the Barnwell Nuclear Fuel Plant (BNFP), the reference fuel reprocessing plant in this report, were grouped into the five categories of hulls and hardware, failed equipment, filters, fluorinator solids, and general process trash (GPT) and sample and analytical cell (SAC) wastes. Six potential treatment options were selected for the five categories of waste. These options represent six basic treatment objectives: (1) no treatment, (2) minimum treatment (compaction), (3) minimum number of processes and products (cementing or grouting), (4) maximum volume reduction without decontamination (melting, incinerating, hot pressing), (5) maximum volume reduction with decontamination (decontamination, treatment of residues), and (6) noncombustible waste forms (melting, incinerating, cementing). Schemes for treatment of each waste type were selected and developed for each treatment option and each type of waste. From these schemes, transuranic waste volumes were found to vary from 1 m/sup 3//MTU for no treatment to as low as 0.02 m/sup 3//MTU. Based on conceptual design requirements, life-cycle costs were estimated for treatment plus on-site storage, transportation, and disposal of both high-level and transuranic wastes (and incremental low-level wastes) from 70,000 MTU. The study concludes that extensive treatment is warranted from both cost and waste form characteristics considerations, and that the characteristics of most of the processing systems used are acceptable. The study recommends that additional combinations of treatment methods or strategies be evaluated and that in the interim, melting, incineration, and cementing be further developed for commercial TRUW. 45 refs., 9 figs., 32 tabs.

  19. Remaining Sites Verification Package for the 100-F-26:14, 116-F-5 Influent Pipelines, Waste Site Reclassification Form 2007-029

    SciTech Connect (OSTI)

    L. M. Dittmer

    2008-02-29

    The 100-F-26:14 waste site includes underground pipelines associated with the 116-F-5 Ball Washer Crib and remnants of process pipelines on the west side of the 105-F Building. 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.

  20. Remaining Sites Verification Package for the 100-B-22:1 Pipelines and Associated Soils, Waste Site Reclassification Form 2005-042

    SciTech Connect (OSTI)

    L. M. Dittmer

    2006-09-12

    The 100-B-22:1 pipelines and associated soils were part of the 100-B Area water treatment facilities. The 100-B-22:1 waste site is limited to those pipelines that interconnected the 185-B Filter House, the 126-B-2 Clearwells, the 185-B Deaeration Plant, and the 190-B Process Pumphouse. None of the 100-B-22:1 pipelines carried environmentally significant contamination. In accordance with the historical information and field observations of this evaluation, the results support a reclassification of this site to No Action required to meet future rural-residential uses and be protective of groundwater and the Columbia River.

  1. Remaining Sites Verification Package for the 141-C Large Animal Barn and Biology Laboratory (Hog Barn), Waste Site Reclassification Form 2006-027

    SciTech Connect (OSTI)

    R. A. Carlson

    2006-05-24

    The 141-C waste site is a former large animal barn and biology laboratory within the 100-F Area experimental animal farm. Strontium-90, arsenic, and multiple polycyclic aromatic hydrocarbons were detected within residual demolition debris at concentrations exceeding cleanup criteria. The site has been remediated by removing approximately 900 bank cubic meters of soil and debris within the former building footprint to the Environmental Restoration Disposal Facility. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  2. Remaining Sites Verification Package for the 126-B-3, 184-B Coal Pit Dumping Area, Waste Site Reclassification Form 2005-028

    SciTech Connect (OSTI)

    L. M. Dittmer

    2006-08-07

    The 126-B-3 waste site is the former coal storage pit for the 184-B Powerhouse. During demolition operations in the 1970s, the site was used for disposal of demolition debris from 100-B/C Area facilities. The site has been remediated by removing debris and contaminated soils. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  3. Canister arrangement for storing radioactive waste

    DOE Patents [OSTI]

    Lorenzo, D.K.; Van Cleve, J.E. Jr.

    1980-04-23

    The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

  4. Bubblers Speed Nuclear Waste Processing at SRS

    SciTech Connect (OSTI)

    2010-11-14

    At the Department of Energy's Savannah River Site, American Recovery and Reinvestment Act funding has supported installation of bubbler technology and related enhancements in the Defense Waste Processing Facility (DWPF). The improvements will accelerate the processing of radioactive waste into a safe, stable form for storage and permit expedited closure of underground waste tanks holding 37 million gallons of liquid nuclear waste.

  5. Canister arrangement for storing radioactive waste

    DOE Patents [OSTI]

    Lorenzo, Donald K. (Knoxville, TN); Van Cleve, Jr., John E. (Kingston, TN)

    1982-01-01

    The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

  6. Bubblers Speed Nuclear Waste Processing at SRS

    ScienceCinema (OSTI)

    None

    2014-08-06

    At the Department of Energy's Savannah River Site, American Recovery and Reinvestment Act funding has supported installation of bubbler technology and related enhancements in the Defense Waste Processing Facility (DWPF). The improvements will accelerate the processing of radioactive waste into a safe, stable form for storage and permit expedited closure of underground waste tanks holding 37 million gallons of liquid nuclear waste.

  7. TRU waste characterization chamber gloveboxes.

    SciTech Connect (OSTI)

    Duncan, D. S.

    1998-07-02

    Argonne National Laboratory-West (ANL-W) is participating in the Department of Energy's (DOE) National Transuranic Waste Program in support of the Waste Isolation Pilot Plant (WIPP). The Laboratory's support currently consists of intrusive characterization of a selected population of drums containing transuranic waste. This characterization is performed in a complex of alpha containment gloveboxes termed the Waste Characterization Gloveboxes. Made up of the Waste Characterization Chamber, Sample Preparation Glovebox, and the Equipment Repair Glovebox, they were designed as a small production characterization facility for support of the Idaho National Engineering and Environmental Laboratory (INEEL). This paper presents salient features of these gloveboxes.

  8. Remaining Sites Verification Package for the 1607-F1 Sanitary Sewer System (124-F-1) and the 100-F-26:8 (1607-F1) Sanitary Sewer Pipelines Waste Sites, Waste Site Reclassification Form 2004-130

    SciTech Connect (OSTI)

    L. M. Dittmer

    2008-03-14

    The 1607-F1 Sanitary Sewer System (124-F-1), consisted of a septic tank, drain field, and associated pipelines that received sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office via the 100-F-26:8 pipelines. The septic tank required remedial action based on confirmatory sampling. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  9. Colloidal agglomerates in tank sludge: Impact on waste processing

    SciTech Connect (OSTI)

    Bunker, B.C.; Martin, J.E.

    1998-06-01

    'Insoluble colloidal sludges in hazardous waste streams such as tank wastes can pose serious problems for waste processing, interfering with retrieval, transport, separation, and solidification procedures. Properties of sediment layers and sludge suspensions such as slurry viscosities, sedimentation rates, and final sediment densities can vary by orders of magnitude depending on the particle types present, the degree to which the particles agglomerate or stick to each other, and on a wide range of processing parameters such as solution shear rates, pH, salt content, and temperature. The objectives of this work are to: (1) understand the factors controlling the nature and extent of colloidal agglomeration under expected waste processing conditions; (2) determine how agglomeration phenomena influence physical properties relevant to waste processing including rheology, sedimentation, and filtration; and (3) develop strategies for optimizing processing conditions via control of agglomeration phenomena. Insoluble colloidal sludges in hazardous waste streams such as tank wastes can pose serious problems for waste processing, interfering with retrieval, transport, separation, and solidification procedures. Properties of sediment layers and sludge suspensions such as slurry viscosities, sedimentation rates, and final sediment densities can vary by orders of magnitude depending on the particle types present, the degree to which the particles agglomerate or stick to each other, and on a wide range of processing parameters such as solution shear rates, pH, salt content, and temperature. The objectives of this work are to: (1) understand the factors controlling the nature and extent of colloidal agglomeration under expected waste processing conditions; (2) determine how agglomeration phenomena influence physical properties relevant to waste processing including rheology, sedimentation, and filtration; and (3) develop strategies for optimizing processing conditions via control of agglomeration phenomena. This project summarizes work performed after almost two years of a three year project. Significant findings include: Particles in Actual Tank Wastes - Transmission electron microscopy of actual wastes shows that most sludges consist of agglomerates of submicron (< 10 -6 m) primary particles of hydrated oxides and insoluble salts. Model colloid suspensions for this work were selected to duplicate the compositions and particle morphologies in actual waste. Agglomeration of Primary Particles - Static light scattering measurements on both model suspensions and actual wastes show that in the basic salt solutions found in most tank wastes, primary particles undergo extensive aggregation to form fractal agglomerates. The fractal nature of the agglomerates has an enormous impact on slurry properties because fractal objects occupy much more space than dense objects at the same solids loading.'

  10. Remaining Sites Verification Package for the 100-F-26:12, 1.8-m (72-in.) Main Process Sewer Pipeline, Waste Site Reclassification Form 2007-034

    SciTech Connect (OSTI)

    J. M. Capron

    2008-04-29

    The 100-F-26:12 waste site was an approximately 308-m-long, 1.8-m-diameter east-west-trending reinforced concrete pipe that joined the North Process Sewer Pipelines (100-F-26:1) and the South Process Pipelines (100-F-26:4) with the 1.8-m reactor cooling water effluent pipeline (100-F-19). In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  11. Remaining Sites Verification Package for the 100-F-44:2, Discovery Pipeline Near 108-F Building, Waste Site Reclassification Form 2007-006

    SciTech Connect (OSTI)

    J. M. Capron

    2008-05-30

    The 100-F-44:2 waste site is a steel pipeline that was discovered in a junction box during confirmatory sampling of the 100-F-26:4 pipeline from December 2004 through January 2005. The 100-F-44:2 pipeline feeds into the 100-F-26:4 subsite vitrified clay pipe (VCP) process sewer pipeline from the 108-F Biology Laboratory at the junction box. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  12. Remaining Sites Verification Package for the 100-B-18, 184-B Powerhouse Debris Pile, Waste Site Reclassification Form 2007-020

    SciTech Connect (OSTI)

    L. M. Dittmer

    2007-11-30

    The 100-B-18 Powerhouse Debris Pile contained miscellaneous demolition waste from the decommissioning activities of the 184-B Powerhouse. The debris covered an area roughly 15 m by 30 m and included materials such as concrete blocks, mixed aggregate/concrete slabs, stone rubble, asphalt rubble, traces of tar/coal, broken fluorescent lights, brick chimney remnants, and rubber hoses. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  13. Conversion of Waste Biomass into Useful Products 

    E-Print Network [OSTI]

    Holtzapple, M.

    1998-01-01

    Waste biomass includes municipal solid waste (MSW), municipal sewage sludge (SS), industrial biosludge, manure, and agricultural residues. When treated with lime, biomass is highly digestible by a mixed culture of acid-forming microorganisms. Lime...

  14. Environmental evaluation of alternatives for long-term management of Defense high-level radioactive wastes at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Not Available

    1982-09-01

    The U.S. Department of Energy (DOE) is considering the selection of a strategy for the long-term management of the defense high-level wastes at the Idaho Chemical Processing Plant (ICPP). This report describes the environmental impacts of alternative strategies. These alternative strategies include leaving the calcine in its present form at the Idaho National Engineering Laboratory (INEL), or retrieving and modifying the calcine to a more durable waste form and disposing of it either at the INEL or in an offsite repository. This report addresses only the alternatives for a program to manage the high-level waste generated at the ICPP. 24 figures, 60 tables.

  15. Technetium Immobilization Forms Literature Survey

    SciTech Connect (OSTI)

    Westsik, Joseph H.; Cantrell, Kirk J.; Serne, R. Jeffrey; Qafoku, Nikolla

    2014-05-01

    Of the many radionuclides and contaminants in the tank wastes stored at the Hanford site, technetium-99 (99Tc) is one of the most challenging to effectively immobilize in a waste form for ultimate disposal. Within the Hanford Tank Waste Treatment and Immobilization Plant (WTP), the Tc will partition between both the high-level waste (HLW) and low-activity waste (LAW) fractions of the tank waste. The HLW fraction will be converted to a glass waste form in the HLW vitrification facility and the LAW fraction will be converted to another glass waste form in the LAW vitrification facility. In both vitrification facilities, the Tc is incorporated into the glass waste form but a significant fraction of the Tc volatilizes at the high glass-melting temperatures and is captured in the off-gas treatment systems at both facilities. The aqueous off-gas condensate solution containing the volatilized Tc is recycled and is added to the LAW glass melter feed. This recycle process is effective in increasing the loading of Tc in the LAW glass but it also disproportionally increases the sulfur and halides in the LAW melter feed which increases both the amount of LAW glass and either the duration of the LAW vitrification mission or the required supplemental LAW treatment capacity.

  16. Quality assurance plan for Final Waste Forms project in support of the development, demonstration, testing and evaluation efforts associated with the Oak Ridge reservation`s LDR/FFCA compliance

    SciTech Connect (OSTI)

    Gilliam, T.M.; Mattus, C.H.

    1994-07-01

    This quality assurance project plan specifies the data quality objectives for Phase I of the Final Waste Forms Project and defines specific measurements and processes required to achieve those objectives. Although the project is funded by the U.S. Department of Energy (DOE), the ultimate recipient of the results is the U.S. Environmental Protection Agency (EPA). Consequently, relevant quality assurance requirements from both organizations must be met. DOE emphasizes administrative structure to ensure quality; EPA`s primary focus is the reproducibility of the generated data. The ten criteria of DOE Order 5700.6C are addressed in sections of this report, while the format used is that prescribed by EPA for quality assurance project plans.

  17. Defense High Level Waste Disposal Container System Description

    SciTech Connect (OSTI)

    2000-10-12

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials will be selected for the disposal container inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lids will be a barrier made of high-nickel alloy. The defense HLW disposal container interfaces with the emplacement drift environment and the internal waste by transferring heat from the canisters to the external environment and by protecting the canisters and their contents from damage/degradation by the external environment. The disposal container also interfaces with the canisters by limiting access of moderator and oxidizing agents to the waste. A loaded and sealed disposal container (waste package) interfaces with the Emplacement Drift System's emplacement drift waste package supports upon which the waste packages are placed. The disposal container interfaces with the Canister Transfer System, Waste Emplacement /Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement, and retrieval for the disposal container/waste package.

  18. NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA

    SciTech Connect (OSTI)

    U.S. DEPARTMENT OF ENERGY, NATIONAL NUCLEAR SECURITY ADMINISTRATION, NEVADA SITE OFFICE

    2005-07-01

    This document establishes the U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive and mixed waste for disposal. Mixed waste generated within the State of Nevada by NNSA/NSO activities is accepted for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the Nevada Test Site Area 3 and Area 5 Radioactive Waste Management Site for storage or disposal.

  19. YUCCA MOUNTAIN WASTE PACKAGE CLOSURE SYSTEM

    SciTech Connect (OSTI)

    G. Housley; C. Shelton-davis; K. Skinner

    2005-08-26

    The method selected for dealing with spent nuclear fuel in the US is to seal the fuel in waste packages and then to place them in an underground repository at the Yucca Mountain Site in Nevada. This article describes the Waste Package Closure System (WPCS) currently being designed for sealing the waste packages.

  20. Rankine cycle waste heat recovery system

    DOE Patents [OSTI]

    Ernst, Timothy C.; Nelson, Christopher R.

    2015-09-22

    A waste heat recovery (WHR) system connects a working fluid to fluid passages formed in an engine block and/or a cylinder head of an internal combustion engine, forming an engine heat exchanger. The fluid passages are formed near high temperature areas of the engine, subjecting the working fluid to sufficient heat energy to vaporize the working fluid while the working fluid advantageously cools the engine block and/or cylinder head, improving fuel efficiency. The location of the engine heat exchanger downstream from an EGR boiler and upstream from an exhaust heat exchanger provides an optimal position of the engine heat exchanger with respect to the thermodynamic cycle of the WHR system, giving priority to cooling of EGR gas. The configuration of valves in the WHR system provides the ability to select a plurality of parallel flow paths for optimal operation.

  1. PROJECT W-551 DETERMINATION DATA FOR EARLY LAW INTERIM PRETREATMENT SYSTEM SELECTION

    SciTech Connect (OSTI)

    TEDESCHI AR

    2008-08-11

    This report provides the detailed assessment forms and data for selection of the solids separation and cesium separation technology for project W-551, Interim Pretreatment System. This project will provide early pretreated low activity waste feed to the Waste Treatment Plant to allow Waste Treatment Plan Low Activity Waste facility operation prior to construction completion of the Pretreatment and High Level Waste facilities. The candidate solids separations technologies are rotary microfiltration and crossflow filtration, and the candidate cesium separation technologies are fractional crystallization, caustic-side solvent extraction, and ion-exchange using spherical resorcinol-formaldehyde resin. This data was used to prepare a cross-cutting technology summary, reported in RPP-RPT-37740.

  2. Efficacy of a solution-based approach for making sodalite waste forms for an oxide reduction salt utilized in the reprocessing of used uranium oxide fuel

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyáš, Josef; Burns, Carolyne A.

    2015-04-01

    This paper describes the various approaches attempted to make solution-derived sodalite with a LiCl-Li2O oxide reduction salt used to dissolve used uranium oxide fuel so the uranium can be recovered and recycled. The approaches include modified sol-gel and solutionbased synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3- SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2Omore »and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions in the crystalline product (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt but that the incorporation of Li into the sodalite is low.« less

  3. Efficacy of a Solution-Based Approach for Making Sodalite Waste Forms for an Oxide Reduction Salt Utilized in the Reprocessing of Used Uranium Oxide Fuel

    SciTech Connect (OSTI)

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyas, Josef; Burns, Carolyn A.

    2015-04-01

    This paper describes various approaches for making sodalite with a LiCl-Li2O oxide reduction salt used to recover uranium from used oxide fuel. The approaches include sol-gel and solution-based synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3-SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2O and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt.

  4. Aluminum phosphate ceramics for waste storage

    SciTech Connect (OSTI)

    Wagh, Arun; Maloney, Martin D

    2014-06-03

    The present disclosure describes solid waste forms and methods of processing waste. In one particular implementation, the invention provides a method of processing waste that may be particularly suitable for processing hazardous waste. In this method, a waste component is combined with an aluminum oxide and an acidic phosphate component in a slurry. A molar ratio of aluminum to phosphorus in the slurry is greater than one. Water in the slurry may be evaporated while mixing the slurry at a temperature of about 140-200.degree. C. The mixed slurry may be allowed to cure into a solid waste form. This solid waste form includes an anhydrous aluminum phosphate with at least a residual portion of the waste component bound therein.

  5. DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

    SciTech Connect (OSTI)

    G. Radulesscu; J.S. Tang

    2000-06-07

    The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this analysis is to support Site Recommendation reports and to assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the Development Plan ''Design Analysis for the Defense High-Level Waste Disposal Container'' (CRWMS M&O 2000c) with no deviations from the plan.

  6. Networks of recyclable material waste-picker’s cooperatives: An alternative for the solid waste management in the city of Rio de Janeiro

    SciTech Connect (OSTI)

    Tirado-Soto, Magda Martina; Zamberlan, Fabio Luiz

    2013-04-15

    Highlights: ? In the marketing of recyclable materials, the waste-pickers are the least wins. ? It is proposed creating a network of recycling cooperatives to achieve viability. ? The waste-pickers contribute to waste management to the city. - Abstract: The objective of this study is to discuss the role of networks formed of waste-picker cooperatives in ameliorating problems of final disposal of solid waste in the city of Rio de Janeiro, since the city’s main landfill will soon have to close because of exhausted capacity. However, it is estimated that in the city of Rio de Janeiro there are around five thousand waste-pickers working in poor conditions, with lack of physical infrastructure and training, but contributing significantly by diverting solid waste from landfills. According to the Sustainable Development Indicators (IBGE, 2010a,b) in Brazil, recycling rates hover between 45% and 55%. In the municipality of Rio de Janeiro, only 1% of the waste produced is collected selectively by the government (COMLURB, 2010), demonstrating that recycling is mainly performed by waste-pickers. Furthermore, since the recycling market is an oligopsony that requires economies of scale to negotiate directly with industries, the idea of working in networks of cooperatives meets the demands for joint marketing of recyclable materials. Thus, this work presents a method for creating and structuring a network of recycling cooperatives, with prior training for working in networks, so that the expected synergies and joint efforts can lead to concrete results. We intend to demonstrate that it is first essential to strengthen the waste-pickers’ cooperatives in terms of infrastructure, governance and training so that solid waste management can be environmentally, socially and economically sustainable in the city of Rio de Janeiro.

  7. Data summary of municipal solid waste management alternatives

    SciTech Connect (OSTI)

    Not Available

    1992-10-01

    This appendix on Mass Burn Technologies is the first in a series designed to identify, describe and assess the suitability of several currently or potentially available generic technologies for the management of municipal solid waste (MSW). These appendices, which cover eight core thermoconversion, bioconversion and recycling technologies, reflect public domain information gathered from many sources. Representative sources include: professional journal articles, conference proceedings, selected municipality solid waste management plans and subscription technology data bases. The information presented is intended to serve as background information that will facilitate the preparation of the technoeconomic and life cycle mass, energy and environmental analyses that are being developed for each of the technologies. Mass burn has been and continues to be the predominant technology in Europe for the management of MSW. In the United States, the majority of the existing waste-to-energy projects utilize this technology and nearly 90 percent of all currently planned facilities have selected mass burn systems. Mass burning generally refers to the direct feeding and combustion of municipal solid waste in a furnace without any significant waste preprocessing. The only materials typically removed from the waste stream prior to combustion are large bulky objects and potentially hazardous or undesirable wastes. The technology has evolved over the last 100 or so years from simple incineration to the most highly developed and commercially proven process available for both reducing the volume of MSW and for recovering energy in the forms of steam and electricity. In general, mass burn plants are considered to operate reliably with high availability.

  8. Charge-coupled substituted garnets (Y 3–x Ca 0.5x M 0.5x )Fe?O?? (M = Ce, Th): Structure and stability as crystalline nuclear waste forms

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Guo, Xiaofeng; Kukkadapu, Ravi K.; Lanzirotti, Antonio; Newville, Matthew; Engelhard, Mark H.; Sutton, Stephen R.; Navrotsky, Alexandra

    2015-04-20

    The garnet structure has been proposed as a potential crystalline nuclear waste form for accommodation of actinide elements, especially uranium (U). In this study, yttrium iron garnet (YIG) as a model garnet host was studied for the incorporation of U analogs, cerium (Ce) and thorium (Th), incorporated by a charge-coupled substitution with calcium (Ca) for yttrium (Y) in YIG, namely, 2Y³? = Ca²? + M??, where M?? = Ce?? or Th??. Single-phase garnets Y3–xCa0.5xM0.5xFe?O?? (x = 0.1–0.7) were synthesized by the citrate–nitrate combustion method. Ce was confirmed to be tetravalent by X-ray absorption spectroscopy and X-ray photoelectron spectroscopy. X-ray diffractionmore »and ??Fe–Mössbauer spectroscopy indicated that M?? and Ca²? cations are restricted to the c site, and the local environments of both the tetrahedral and the octahedral Fe³? are systematically affected by the extent of substitution. The charge-coupled substitution has advantages in incorporating Ce/Th and in stabilizing the substituted phases compared to a single substitution strategy. Enthalpies of formation of garnets were obtained by high-temperature oxide melt solution calorimetry, and the enthalpies of substitution of Ce and Th were determined. The thermodynamic analysis demonstrates that the substituted garnets are entropically rather than energetically stabilized. This suggests that such garnets may form and persist in repositories at high temperature but might decompose near room temperature.« less

  9. Method of preparing nuclear wastes for tansportation and interim storage

    DOE Patents [OSTI]

    Bandyopadhyay, Gautam (Naperville, IL); Galvin, Thomas M. (Darien, IL)

    1984-01-01

    Nuclear waste is formed into a substantially water-insoluble solid for temporary storage and transportation by mixing the calcined waste with at least 10 weight percent powdered anhydrous sodium silicate to form a mixture and subjecting the mixture to a high humidity environment for a period of time sufficient to form cementitious bonds by chemical reaction. The method is suitable for preparing an interim waste form from dried high level radioactive wastes.

  10. Effect of potential waste constituents on the reactivity of Hanford ferrocyanide wastes: Diluent, catalyst, and initiator studies

    SciTech Connect (OSTI)

    Scheele, R.D.; Johnston, J.W.; Tingey, J.M.; Burger, L.L.; Sell, R.L.

    1993-04-01

    During the 1980s, scientists at the Hanford Site began considering disposal options for wastes in underground storage tanks. As a result of safety concerns, it was determined that special consideration should be given to ferrocyanide-bearing wastes to ensure their continued safe storage. In addition, Westinghouse Hanford Company (WHC) chartered Pacific Northwest Laboratory (PNL) to determine the conditions necessary for vigorous reactions to occur in the Hanford Site ferrocyanide wastes. As part of those studies, PNL has evaluated the effects of selected potential waste constituents to determine how they might affect the reactivity of the wastes. The authors` investigations of the diluent, catalytic, or initiating effects of potential waste constituents included studies (1) to determine the effect of the oxidant-to-ferrocyanide ratio, (2) to establish the effect of sodium aluminate concentration, (3) to identify materials that could affect the explosivity of a mixture of sodium nickel ferricyanide (a potential aging product of ferrocyanide) and sodium nitrate and nitrite, (4) and to determine the effect of nickel sulfide concentration. They also conducted a thermal sensitivity study and analyzed the results to determine the relative behaviors of sodium nickel ferrocyanide and ferricyanide. A statistical evaluation of the time-to-explosion (TTX) test results from the catalyst and initiator screening study found that the ferricyanide reacted at a faster rate than did the ferrocyanide analog. The thermal analyses indicated that the ferricyanide form is more thermally sensitive, exhibiting exothermic behavior at a lower temperature than the ferrocyanide form. The increased thermal sensitivity of the ferricyanide, which is a potential oxidation product of ferrocyanide, relative to the ferrocyanide analog, does not support the hypothesis that aging independent of the reaction pathway will necessarily reduce the reaction hazard of ferrocyanide wastes.

  11. Naval Waste Package Design Report

    SciTech Connect (OSTI)

    M.M. Lewis

    2004-03-15

    A design methodology for the waste packages and ancillary components, viz., the emplacement pallets and drip shields, has been developed to provide designs that satisfy the safety and operational requirements of the Yucca Mountain Project. This methodology is described in the ''Waste Package Design Methodology Report'' Mecham 2004 [DIRS 166168]. To demonstrate the practicability of this design methodology, four waste package design configurations have been selected to illustrate the application of the methodology. These four design configurations are the 21-pressurized water reactor (PWR) Absorber Plate waste package, the 44-boiling water reactor (BWR) waste package, the 5-defense high-level waste (DHLW)/United States (U.S.) Department of Energy (DOE) spent nuclear fuel (SNF) Co-disposal Short waste package, and the Naval Canistered SNF Long waste package. Also included in this demonstration is the emplacement pallet and continuous drip shield. The purpose of this report is to document how that design methodology has been applied to the waste package design configurations intended to accommodate naval canistered SNF. This demonstrates that the design methodology can be applied successfully to this waste package design configuration and support the License Application for construction of the repository.

  12. Waste management system alternatives for treatment of wastes from spent fuel reprocessing

    SciTech Connect (OSTI)

    McKee, R.W.; Swanson, J.L.; Daling, P.M.; Clark, L.L.; Craig, R.A.; Nesbitt, J.F.; McCarthy, D.; Franklin, A.L.; Hazelton, R.F.; Lundgren, R.A.

    1986-09-01

    This study was performed to help identify a preferred TRU waste treatment alternative for reprocessing wastes with respect to waste form performance in a geologic repository, near-term waste management system risks, and minimum waste management system costs. The results were intended for use in developing TRU waste acceptance requirements that may be needed to meet regulatory requirements for disposal of TRU wastes in a geologic repository. The waste management system components included in this analysis are waste treatment and packaging, transportation, and disposal. The major features of the TRU waste treatment alternatives examined here include: (1) packaging (as-produced) without treatment (PWOT); (2) compaction of hulls and other compactable wastes; (3) incineration of combustibles with cementation of the ash plus compaction of hulls and filters; (4) melting of hulls and failed equipment plus incineration of combustibles with vitrification of the ash along with the HLW; (5a) decontamination of hulls and failed equipment to produce LLW plus incineration and incorporation of ash and other inert wastes into HLW glass; and (5b) variation of this fifth treatment alternative in which the incineration ash is incorporated into a separate TRU waste glass. The six alternative processing system concepts provide progressively increasing levels of TRU waste consolidation and TRU waste form integrity. Vitrification of HLW and intermediate-level liquid wastes (ILLW) was assumed in all cases.

  13. Hanford Tank Waste - Near Source Treatment of Low Activity Waste

    SciTech Connect (OSTI)

    Ramsey, William Gene

    2013-08-15

    Abstract only. Treatment and disposition of Hanford Site waste as currently planned consists of 100+ waste retrievals, waste delivery through up to 8+ miles of dedicated, in-ground piping, centralized mixing and blending operations- all leading to pre-treatment combination and separation processes followed by vitrification at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The sequential nature of Tank Farm and WTP operations requires nominally 15-20 years of continuous operations before all waste can be retrieved from many Single Shell Tanks (SSTs). Also, the infrastructure necessary to mobilize and deliver the waste requires significant investment beyond that required for the WTP. Treating waste as closely as possible to individual tanks or groups- as allowed by the waste characteristics- is being investigated to determine the potential to 1) defer, reduce, and/or eliminate infrastructure requirements, and 2) significantly mitigate project risk by reducing the potential and impact of single point failures. The inventory of Hanford waste slated for processing and disposition as LAW is currently managed as high-level waste (HLW), i.e., the separation of fission products and other radionuclides has not commenced. A significant inventory of this waste (over 20M gallons) is in the form of precipitated saltcake maintained in single shell tanks, many of which are identified as potential leaking tanks. Retrieval and transport (as a liquid) must be staged within the waste feed delivery capability established by site infrastructure and WTP. Near Source treatment, if employed, would provide for the separation and stabilization processing necessary for waste located in remote farms (wherein most of the leaking tanks reside) significantly earlier than currently projected. Near Source treatment is intended to address the currently accepted site risk and also provides means to mitigate future issues likely to be faced over the coming decades. This paper describes the potential near source treatment and waste disposition options as well as the impact these options could have on reducing infrastructure requirements, project cost and mission schedule.

  14. Aqueous Zinc Bromide Waste Solidification

    SciTech Connect (OSTI)

    Langton, C.A.

    2002-07-23

    The goal of this study was to select one or more commercially available aqueous sorbents to solidify the zinc bromide solution stored in C-Area, identify the polymer to zinc bromide solution ratio (waste loading) for the selected sorbents, and identify processing issues that require further testing in pilot-scale testing.

  15. Los Alamos National Laboratory Hazardous Waste Facility Permit...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Hazardous Waste Facility Permit Draft Community Relations Plan CommentSuggestion Form Instructions for completing the form: Please reference the section in the plan that your...

  16. Nevada Test Site Waste Acceptance Criteria

    SciTech Connect (OSTI)

    U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office

    2005-10-01

    This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive (LLW) and mixed waste (MW) for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex (RWMC) for storage or disposal.

  17. Engineered waste-package-system design specification

    SciTech Connect (OSTI)

    Not Available

    1983-05-01

    This report documents the waste package performance requirements and geologic and waste form data bases used in developing the conceptual designs for waste packages for salt, tuff, and basalt geologies. The data base reflects the latest geotechnical information on the geologic media of interest. The parameters or characteristics specified primarily cover spent fuel, defense high-level waste, and commercial high-level waste forms. The specification documents the direction taken during the conceptual design activity. A separate design specification will be developed prior to the start of the preliminary design activity.

  18. The Waste Isolation Pilot Plant Hazardous Waste Facility Permit...

    Office of Environmental Management (EM)

    The Waste Isolation Pilot Plant Hazardous Waste Facility Permit, Waste Analysis Plan The Waste Isolation Pilot Plant Hazardous Waste Facility Permit, Waste Analysis Plan This...

  19. Radioactive waste material melter apparatus

    DOE Patents [OSTI]

    Newman, D.F.; Ross, W.A.

    1990-04-24

    An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs.

  20. Radioactive waste material melter apparatus

    DOE Patents [OSTI]

    Newman, Darrell F. (Richland, WA); Ross, Wayne A. (Richland, WA)

    1990-01-01

    An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another.

  1. Process for treating fission waste. [Patent application

    DOE Patents [OSTI]

    Rohrmann, C.A.; Wick, O.J.

    1981-11-17

    A method is described for the treatment of fission waste. A glass forming agent, a metal oxide, and a reducing agent are mixed with the fission waste and the mixture is heated. After melting, the mixture separates into a glass phase and a metal phase. The glass phase may be used to safely store the fission waste, while the metal phase contains noble metals recovered from the fission waste.

  2. Method for processing aqueous wastes

    DOE Patents [OSTI]

    Pickett, J.B.; Martin, H.L.; Langton, C.A.; Harley, W.W.

    1993-12-28

    A method is presented for treating waste water such as that from an industrial processing facility comprising the separation of the waste water into a dilute waste stream and a concentrated waste stream. The concentrated waste stream is treated chemically to enhance precipitation and then allowed to separate into a sludge and a supernate. The supernate is skimmed or filtered from the sludge and blended with the dilute waste stream to form a second dilute waste stream. The sludge remaining is mixed with cementitious material, rinsed to dissolve soluble components, then pressed to remove excess water and dissolved solids before being allowed to cure. The dilute waste stream is also chemically treated to decompose carbonate complexes and metal ions and then mixed with cationic polymer to cause the precipitated solids to flocculate. Filtration of the flocculant removes sufficient solids to allow the waste water to be discharged to the surface of a stream. The filtered material is added to the sludge of the concentrated waste stream. The method is also applicable to the treatment and removal of soluble uranium from aqueous streams, such that the treated stream may be used as a potable water supply. 4 figures.

  3. Method for processing aqueous wastes

    DOE Patents [OSTI]

    Pickett, John B. (3922 Wood Valley Dr., Aiken, SC 29803); Martin, Hollis L. (Rt. 1, Box 188KB, McCormick, SC 29835); Langton, Christine A. (455 Sumter St. SE., Aiken, SC 29801); Harley, Willie W. (110 Fairchild St., Batesburg, SC 29006)

    1993-01-01

    A method for treating waste water such as that from an industrial processing facility comprising the separation of the waste water into a dilute waste stream and a concentrated waste stream. The concentrated waste stream is treated chemically to enhance precipitation and then allowed to separate into a sludge and a supernate. The supernate is skimmed or filtered from the sludge and blended with the dilute waste stream to form a second dilute waste stream. The sludge remaining is mixed with cementitious material, rinsed to dissolve soluble components, then pressed to remove excess water and dissolved solids before being allowed to cure. The dilute waste stream is also chemically treated to decompose carbonate complexes and metal ions and then mixed with cationic polymer to cause the precipitated solids to flocculate. Filtration of the flocculant removes sufficient solids to allow the waste water to be discharged to the surface of a stream. The filtered material is added to the sludge of the concentrated waste stream. The method is also applicable to the treatment and removal of soluble uranium from aqueous streams, such that the treated stream may be used as a potable water supply.

  4. Geological challenges in radioactive waste isolation: Third worldwide review

    E-Print Network [OSTI]

    Witherspoon editor, P.A.; Bodvarsson editor, G.S.

    2001-01-01

    International Atomic Energy Agency (IAEA), Underground Disposal of Radioactive Waste,International Atomic Energy Agency (IAEA) regulations. The site- selection procedure began with exploring the possibility of radioactive waste

  5. Spectroelectrochemical Sensing Based on Multimode Selectivity Simultaneously Achievable in a Single Device 10. Sensing of Ferrocyanide in Hanford Waste Simulant Solution

    SciTech Connect (OSTI)

    Maizels, Mila; Seliskar, Carl J.; Heineman, William R.; Bryan, Samuel A.

    2002-11-01

    A ferrocyanide spectroelectrochemical sensor has been evaluated on Hanford U-Plant-2 simulant solution. The sensor consists of an indium tin oxide optically transparent electrode coated with a thin film of porous silica containing immobilized PDMDAAC as an anion exchanger. Ferrocyanide is quantified by the absorbance change at 420 nm that accompanies electrochemical modulation of ferrocyanide that partitions into the film. Hanford U-Plant-2 simulant mimics the contents of tanks in which radioactive waste is stored. A solution of the simulant is a complex mixture of high pH (10) and high ionic strength that contains ferrocyanide. Systematic evaluation of the species in the simulant solution showed the only significant interference to be competition by the anions in the simulant with the preconcentration of ferrocyanide into the sensing film. Deterioration of the sensing film by the high ionic strength and alkaline conditions restricted its useful lifetime to about 40 min. Analysis of simulant with a sensor calibrated at high ionic strength gave 8.0 mM ferrocyanide compared to 8.2 mM obtained by FTIR.

  6. International low level waste disposal practices and facilities

    SciTech Connect (OSTI)

    Nutt, W.M.

    2011-12-19

    The safe management of nuclear waste arising from nuclear activities is an issue of great importance for the protection of human health and the environment now and in the future. The primary goal of this report is to identify the current situation and practices being utilized across the globe to manage and store low and intermediate level radioactive waste. The countries included in this report were selected based on their nuclear power capabilities and involvement in the nuclear fuel cycle. This report highlights the nuclear waste management laws and regulations, current disposal practices, and future plans for facilities of the selected international nuclear countries. For each country presented, background information and the history of nuclear facilities are also summarized to frame the country's nuclear activities and set stage for the management practices employed. The production of nuclear energy, including all the steps in the nuclear fuel cycle, results in the generation of radioactive waste. However, radioactive waste may also be generated by other activities such as medical, laboratory, research institution, or industrial use of radioisotopes and sealed radiation sources, defense and weapons programs, and processing (mostly large scale) of mineral ores or other materials containing naturally occurring radionuclides. Radioactive waste also arises from intervention activities, which are necessary after accidents or to remediate areas affected by past practices. The radioactive waste generated arises in a wide range of physical, chemical, and radiological forms. It may be solid, liquid, or gaseous. Levels of activity concentration can vary from extremely high, such as levels associated with spent fuel and residues from fuel reprocessing, to very low, for instance those associated with radioisotope applications. Equally broad is the spectrum of half-lives of the radionuclides contained in the waste. These differences result in an equally wide variety of options for the management of radioactive waste. There is a variety of alternatives for processing waste and for short term or long term storage prior to disposal. Likewise, there are various alternatives currently in use across the globe for the safe disposal of waste, ranging from near surface to geological disposal, depending on the specific classification of the waste. At present, there appears to be a clear and unequivocal understanding that each country is ethically and legally responsible for its own wastes, in accordance with the provisions of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. Therefore the default position is that all nuclear wastes will be disposed of in each of the 40 or so countries concerned with nuclear power generation or part of the fuel cycle. To illustrate the global distribution of radioactive waste now and in the near future, Table 1 provides the regional breakdown, based on the UN classification of the world in regions illustrated in Figure 1, of nuclear power reactors in operation and under construction worldwide. In summary, 31 countries operate 433 plants, with a total capacity of more than 365 gigawatts of electrical energy (GW[e]). A further 65 units, totaling nearly 63 GW(e), are under construction across 15 of these nations. In addition, 65 countries are expressing new interest in, considering, or actively planning for nuclear power to help address growing energy demands to fuel economic growth and development, climate change concerns, and volatile fossil fuel prices. Of these 65 new countries, 21 are in Asia and the Pacific region, 21 are from the Africa region, 12 are in Europe (mostly Eastern Europe), and 11 in Central and South America. However, 31 of these 65 are not currently planning to build reactors, and 17 of those 31 have grids of less than 5 GW, which is said to be too small to accommodate most of the reactor designs available. For the remaining 34 countries actively planning reactors, as of September 2010: 14 indicate a strong intention to precede w

  7. Method of waste stabilization with dewatered chemically bonded phosphate ceramics

    DOE Patents [OSTI]

    Wagh, Arun; Maloney, Martin D.

    2010-06-29

    A method of stabilizing a waste in a chemically bonded phosphate ceramic (CBPC). The method consists of preparing a slurry including the waste, water, an oxide binder, and a phosphate binder. The slurry is then allowed to cure to a solid, hydrated CBPC matrix. Next, bound water within the solid, hydrated CBPC matrix is removed. Typically, the bound water is removed by applying heat to the cured CBPC matrix. Preferably, the quantity of heat applied to the cured CBPC matrix is sufficient to drive off water bound within the hydrated CBPC matrix, but not to volatalize other non-water components of the matrix, such as metals and radioactive components. Typically, a temperature range of between 100.degree. C.-200.degree. C. will be sufficient. In another embodiment of the invention wherein the waste and water have been mixed prior to the preparation of the slurry, a select amount of water may be evaporated from the waste and water mixture prior to preparation of the slurry. Another aspect of the invention is a direct anyhydrous CBPC fabrication method wherein water is removed from the slurry by heating and mixing the slurry while allowing the slurry to cure. Additional aspects of the invention are ceramic matrix waste forms prepared by the methods disclosed above.

  8. MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION FINAL REPORT 08R1360-1

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT W; PEGG IL; JOSEPH I; BARDAKCI T; GAN H; GONG W; CHAUDHURI M

    2010-01-04

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in wasteloading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  9. State of the art review of alternatives to shallow land burial of low level radioactive waste

    SciTech Connect (OSTI)

    Not Available

    1980-04-01

    A review of alternatives to shallow land burial for disposal of low level radioactive waste was conducted to assist ORNL in developing a program for the evaluation, selection, and demonstration of the most acceptable alternatives. The alternatives were categorized as follows: (1) near term isolation concepts, (2) far term isolation concepts, (3) dispersion concepts, and (4) conversion concepts. Detailed descriptions of near term isolation concepts are provided. The descriptions include: (1) method of isolation, (2) waste forms that can be accommodated, (3) advantages and disadvantages, (4) facility and equipment requirements, (5) unusual operational or maintenance requirements, (6) information/technology development requirements, and (7) related investigations of the concept.

  10. Tank Waste Remediation System optimized processing strategy

    SciTech Connect (OSTI)

    Slaathaug, E.J.; Boldt, A.L.; Boomer, K.D.; Galbraith, J.D.; Leach, C.E.; Waldo, T.L.

    1996-03-01

    This report provides an alternative strategy evolved from the current Hanford Site Tank Waste Remediation System (TWRS) programmatic baseline for accomplishing the treatment and disposal of the Hanford Site tank wastes. This optimized processing strategy performs the major elements of the TWRS Program, but modifies the deployment of selected treatment technologies to reduce the program cost. The present program for development of waste retrieval, pretreatment, and vitrification technologies continues, but the optimized processing strategy reuses a single facility to accomplish the separations/low-activity waste (LAW) vitrification and the high-level waste (HLW) vitrification processes sequentially, thereby eliminating the need for a separate HLW vitrification facility.

  11. MUSHROOM WASTE MANAGEMENT PROJECT LIQUID WASTE MANAGEMENT

    E-Print Network [OSTI]

    #12;MUSHROOM WASTE MANAGEMENT PROJECT LIQUID WASTE MANAGEMENT PHASE I: AUDIT OF CURRENT PRACTICE The Mushroom Waste Management Project (MWMP) was initiated by Environment Canada, the BC Ministry of solid and liquid wastes generated at mushroom producing facilities. Environmental guidelines

  12. RCRA information on hazardous wastes for publicly owned treatment works. Technical report

    SciTech Connect (OSTI)

    Not Available

    1985-09-01

    The guidance manual provides guidance to municipal personnel in understanding hazardous waste requirements of the Resource Conservation and Recovery Act (RCRA) and the implications of these RCRA requirements for the wastewater treatment plant operated by your municipality, for your local pretreatment program, and for local industries served by the treatment plant. The primary purpose of the manual is the RCRA notification requirement specified in the General Pretreatment Regulations. The manual focuses on Subtitle C requirements. (Subtitle C is directly applicable to industries since this program regulates generators, transporters, and disposers of hazardous waste). The manual also provides a general understanding of how federal RCRA requirements for hazardous waste affect industrial users. The manual also will be helpful in complying with any applicable federal requirements incumbent upon your POTW under Subtitle C of RCRA. The appendices contain lists of hazardous wastes regulated by federal requirements; selected EPA-approved forms for hazardous waste facilities to use; RCRA information brochure which briefly outlines the Act's impact on industries that generate or transport hazardous wastes; and EPA pamphlets summarizing information for generators of small quantities of hazardous waste.

  13. Waste management plan for Hanford spent nuclear fuel characterization activities

    SciTech Connect (OSTI)

    Chastain, S.A. [Westinghouse Hanford Co., Richland, WA (United States); Spinks, R.L. [Pacific Northwest Lab., Richland, WA (United States)

    1994-10-17

    A joint project was initiated between Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratory (PNL) to address critical issues associated with the Spent Nuclear Fuel (SNF) stored at the Hanford Site. Recently, particular attention has been given to remediation of the SNF stored in the K Basins. A waste management plan (WMP) acceptable to both parties is required prior to the movement of selected material to the PNL facilities for examination. N Reactor and Single Pass Reactor (SPR) fuel has been stored for an extended period of time in the N Reactor, PUREX, K-East, and K-West Basins. Characterization plans call for transport of fuel material form the K Basins to the 327 Building Postirradiation Testing Laboratory (PTL) in the 300 Area for examination. However, PNL received a directive stating that no examination work will be started in PNL hot cell laboratories without an approved disposal route for all waste generated related to the activity. Thus, as part of the Characterization Program Management Plan for Hanford Spent Nuclear Fuel, a waste management plan which will ensure that wastes generated as a result of characterization activities conducted at PNL will be accepted by WHC for disposition is required. This document contains the details of the waste handling plan that utilizes, to the greatest extent possible, established waste handling and disposal practices at Hanford between PNL and WHC. Standard practices are sufficient to provides for disposal of most of the waste materials, however, special consideration must be given to the remnants of spent nuclear fuel elements following examination. Fuel element remnants will be repackaged in an acceptable container such as the single element canister and returned to the K Basins for storage.

  14. Municipal waste processing apparatus

    DOE Patents [OSTI]

    Mayberry, John L. (Idaho Falls, ID)

    1988-01-01

    Municipal waste materials are processed by crushing the materials so that pieces of noncombustible material are smaller than a selected size and pieces of combustible material are larger than the selected size. The crushed materials are placed on a vibrating mesh screen conveyor belt having openings which pass the smaller, noncombustible pieces of material, but do not pass the larger, combustible pieces of material. Pieces of material which become lodged in the openings of the conveyor belt may be removed by cylindrical deraggers or pressurized air. The crushed materials may be fed onto the conveyor belt by a vibrating feed plate which shakes the materials so that they tend to lie flat.

  15. Municipal waste processing apparatus

    DOE Patents [OSTI]

    Mayberry, John L. (Idaho Falls, ID)

    1989-01-01

    Municipal waste materials are processed by crushing the materials so that pieces of noncombustible material are smaller than a selected size and pieces of combustible material are larger than the selected size. The crushed materials are placed on a vibrating mesh screen conveyor belt having openings which pass the smaller, noncombustible pieces of material, but do not pass the larger, combustible pieces of material. Consecutive conveyors may be connected by an intermediate vibratory plate. An air knife can be used to further separate materials based on weight.

  16. Nevada National Security Site Waste Acceptance Criteria

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2011-01-01

    This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

  17. Nevada National Security Site Waste Acceptance Criteria

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2010-09-03

    This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

  18. Deep Borehole Disposal Research: Demonstration Site Selection...

    Energy Savers [EERE]

    radioactive waste forms, along with research and development for mined repositories in salt, granite, and clay, as part of the used fuel disposition (UFD) campaign. The deep...

  19. Waste drum gas generation sampling program at Rocky Flats during FY 1989

    SciTech Connect (OSTI)

    Roggenthen, D.K.; Nieweg, R.G.

    1990-10-01

    Rocky Flats Plant transuranic waste drums were sampled for gas composition. Glass, metal, graphite, and solidified inorganic sludge transuranic waste forms were sampled. A vacuum system was used to sample each layer of containment inside a waste drum, including individual waste bags. G values were calculated for the waste drums. G(H{sub 2}) was below 0.6 and G(Total) was below 1.3 for all waste forms discussed in this report. 5 refs., 3 figs., 3 tabs.

  20. Radioactive waste treatment technologies and environment

    SciTech Connect (OSTI)

    HORVATH, Jan; KRASNY, Dusan [JAVYS, PLc. - Nuclear and Decommisioning Company, PLc. (Slovakia)

    2007-07-01

    The radioactive waste treatment and conditioning are the most important steps in radioactive waste management. At the Slovak Electric, plc, a range of technologies are used for the processing of radioactive waste into a form suitable for disposal in near surface repository. These technologies operated by JAVYS, PLc. Nuclear and Decommissioning Company, PLc. Jaslovske Bohunice are described. Main accent is given to the Bohunice Radwaste Treatment and Conditioning Centre, Bituminization plant, Vitrification plant, and Near surface repository of radioactive waste in Mochovce and their operation. Conclusions to safe and effective management of radioactive waste in the Slovak Republic are presented. (authors)

  1. Integrated solid waste management of Scottsdale, Arizona

    SciTech Connect (OSTI)

    1995-11-01

    The subject document reports the results of an in-depth investigation of the fiscal year 1992 cost of the city of Scottsdale, Arizona, integrated municipal solid waste management (IMSWM) system, the energy consumed to operate the system, and the environmental performance requirements for each of the system`s waste-processing and disposal facilities. The document reports actual data from records kept by participants. Every effort was made to minimize the use of assumptions, and no attempt is made to interpret the data reported. Analytical approaches are documented so that interested analysts may per-form manipulation or further analysis of the data. As such, the report is a reference document for municipal solid waste (MSW) management professionals who are interested in the actual costs and energy consumption, for a 1-year period, of an operating IMSWM system. The report is organized into two main parts. The first part is the executive summary and case study portion of the report. The executive summary provides a basic description of the study area and selected economic and energy information. Within the case study are detailed descriptions of each component operating during the study period; the quantities of solid waste collected, processed, and marketed within the study boundaries; the cost of MSW in Scottsdale; an energy usage analysis; a review of federal, state, and local environmental requirement compliance; a reference section; and a glossary of terms. The second part of the report focuses on a more detailed discourse on the above topics. In addition, the methodology used to determine the economic costs and energy consumption of the system components is found in the second portion of this report. The methodology created for this project will be helpful for those professionals who wish to break out the costs of their own integrated systems.

  2. SECONDARY WASTE MANAGEMENT FOR HANFORD EARLY LOW ACTIVITY WASTE VITRIFICATION

    SciTech Connect (OSTI)

    UNTERREINER BJ

    2008-07-18

    More than 200 million liters (53 million gallons) of highly radioactive and hazardous waste is stored at the U.S. Department of Energy's Hanford Site in southeastern Washington State. The DOE's Hanford Site River Protection Project (RPP) mission includes tank waste retrieval, waste treatment, waste disposal, and tank farms closure activities. This mission will largely be accomplished by the construction and operation of three large treatment facilities at the Waste Treatment and Immobilization Plant (WTP): (1) a Pretreatment (PT) facility intended to separate the tank waste into High Level Waste (HLW) and Low Activity Waste (LAW); (2) a HLW vitrification facility intended to immobilize the HLW for disposal at a geologic repository in Yucca Mountain; and (3) a LAW vitrification facility intended to immobilize the LAW for shallow land burial at Hanford's Integrated Disposal Facility (IDF). The LAW facility is on target to be completed in 2014, five years prior to the completion of the rest of the WTP. In order to gain experience in the operation of the LAW vitrification facility, accelerate retrieval from single-shell tank (SST) farms, and hasten the completion of the LAW immobilization, it has been proposed to begin treatment of the low-activity waste five years before the conclusion of the WTP's construction. A challenge with this strategy is that the stream containing the LAW vitrification facility off-gas treatment condensates will not have the option of recycling back to pretreatment, and will instead be treated by the Hanford Effluent Treatment Facility (ETF). Here the off-gas condensates will be immobilized into a secondary waste form; ETF solid waste.

  3. EM Waste Acceptance Product Specification (WAPS) for Vitrified...

    Office of Environmental Management (EM)

    EM Waste Acceptance Product Specification (WAPS) for Vitrified High-Level Waste Forms Presentation to the HLW Corporate Board July 24, 2008 By Tony KlukKen Picha 2 Background *...

  4. Municipal Solid Waste Combustion : Fuel Testing and Characterization : Task 1 Report, May 30, 1990-October 1, 1990.

    SciTech Connect (OSTI)

    Bushnell, Dwight J.; Canova, Joseph H.; Dadkhah-Nikoo, Abbas.

    1990-10-01

    The objective of this study is to screen and characterize potential biomass fuels from waste streams. This will be accomplished by determining the types of pollutants produced while burning selected municipal waste, i.e., commercial mixed waste paper residential (curbside) mixed waste paper, and refuse derived fuel. These materials will be fired alone and in combination with wood, equal parts by weight. The data from these experiments could be utilized to size pollution control equipment required to meet emission standards. This document provides detailed descriptions of the testing methods and evaluation procedures used in the combustion testing and characterization project. The fuel samples will be examined thoroughly from the raw form to the exhaust emissions produced during the combustion test of a densified sample.

  5. Nuclear waste management. Quarterly progress report, July-September 1980

    SciTech Connect (OSTI)

    Chikalla, T.D.

    1980-11-01

    Research is reported on: high-level waste immobilization, alternative waste forms, TRU waste immobilization and decontamination, krypton solidification, thermal outgassing, /sup 129/I fixation, unsaturated zone transport, well-logging instrumentation, waste management system and safety studies, effectiveness of geologic isolation systems, waste/rock interactions, engineered barriers, backfill material, spent fuel storage (criticality), barrier sealing and liners for U mill tailings, and revegetation of inactive U tailings sites. (DLC)

  6. Waste acceptance criteria for the Waste Isolation Pilot Plant

    SciTech Connect (OSTI)

    NONE

    1996-04-01

    The Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC), DOE/WIPP-069, was initially developed by a U.S. Department of Energy (DOE) Steering Committee to provide performance requirements to ensure public health and safety as well as the safe handling of transuranic (TRU) waste at the WIPP. This revision updates the criteria and requirements of previous revisions and deletes those which were applicable only to the test phase. The criteria and requirements in this document must be met by participating DOE TRU Waste Generator/Storage Sites (Sites) prior to shipping contact-handled (CH) and remote-handled (RH) TRU waste forms to the WIPP. The WIPP Project will comply with applicable federal and state regulations and requirements, including those in Titles 10, 40, and 49 of the Code of Federal Regulations (CFR). The WAC, DOE/WIPP-069, serves as the primary directive for assuring the safe handling, transportation, and disposal of TRU wastes in the WIPP and for the certification of these wastes. The WAC identifies strict requirements that must be met by participating Sites before these TRU wastes may be shipped for disposal in the WIPP facility. These criteria and requirements will be reviewed and revised as appropriate, based on new technical or regulatory requirements. The WAC is a controlled document. Revised/changed pages will be supplied to all holders of controlled copies.

  7. Tritium waste package

    DOE Patents [OSTI]

    Rossmassler, R.; Ciebiera, L.; Tulipano, F.J.; Vinson, S.; Walters, R.T.

    1995-11-07

    A containment and waste package system for processing and shipping tritium oxide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within the outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen and oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB. 1 fig.

  8. Hazardous Waste Pick-up DEPARTMENT of ENVIRONMENTAL HEALTH and SAFETY

    E-Print Network [OSTI]

    Emmons, Scott

    Hazardous Waste Pick-up Form DEPARTMENT of ENVIRONMENTAL HEALTH and SAFETY Albert Einstein College: Principal Investigator: Extension: Location of Waste: Email: Waste Description If mixed, list all known Hazardous Waste label. Any container without a Hazardous Waste label cannot be picked up by Environmental

  9. Waste Treatment Plant - 12508

    SciTech Connect (OSTI)

    Harp, Benton; Olds, Erik

    2012-07-01

    The Waste Treatment Plant (WTP) will immobilize millions of gallons of Hanford's tank waste into solid glass using a proven technology called vitrification. The vitrification process will turn the waste into a stable glass form that is safe for long-term storage. Our discussion of the WTP will include a description of the ongoing design and construction of this large, complex, first-of-a-kind project. The concept for the operation of the WTP is to separate high-level and low-activity waste fractions, and immobilize those fractions in glass using vitrification. The WTP includes four major nuclear facilities and various support facilities. Waste from the Tank Farms is first pumped to the Pretreatment Facility at the WTP through an underground pipe-in-pipe system. When construction is complete, the Pretreatment Facility will be 12 stories high, 540 feet long and 215 feet wide, making it the largest of the four major nuclear facilities that compose the WTP. The total size of this facility will be more than 490,000 square feet. More than 8.2 million craft hours are required to construct this facility. Currently, the Pretreatment Facility is 51 percent complete. At the Pretreatment Facility the waste is pumped to the interior waste feed receipt vessels. Each of these four vessels is 55-feet tall and has a 375,000 gallon capacity, which makes them the largest vessels inside the Pretreatment Facility. These vessels contain a series of internal pulse-jet mixers to keep incoming waste properly mixed. The vessels are inside the black-cell areas, completely enclosed behind thick steel-laced, high strength concrete walls. The black cells are designed to be maintenance free with no moving parts. Once hot operations commence the black-cell area will be inaccessible. Surrounded by black cells, is the 'hot cell canyon'. The hot cell contains all the moving and replaceable components to remove solids and extract liquids. In this area, there is ultrafiltration equipment, cesium-ion exchange columns, evaporator boilers and recirculation pumps, and various mechanical process pumps for transferring process fluids. During the first phase of pretreatment, the waste will be concentrated using an evaporation process. Solids will be filtered out, and the remaining soluble, highly radioactive isotopes will be removed using an ion-exchange process. The high-level solids will be sent to the High-Level Waste (HLW) Vitrification Facility, and the low activity liquids will be sent to the Low-Activity Waste (LAW) Vitrification Facility for further processing. The high-level waste will be transferred via underground pipes to the HLW Facility from the Pretreatment Facility. The waste first arrives at the wet cell, which rests inside a black-cell area. The pretreated waste is transferred through shielded pipes into a series of melter preparation and feed vessels before reaching the melters. Liquids from various facility processes also return to the wet cell for interim storage before recycling back to the Pretreatment Facility. (authors)

  10. Selection of Working Fluids for the Organic Rankine Cycle 

    E-Print Network [OSTI]

    West, H. H.; Patton, J. M.; Starling, K. E.

    1979-01-01

    The subject of selecting working fluid and process operating conditions for the waste heat binary power cycle is addressed herein. The waste heat temperature range from 300 F to 500 F was considered the economic resource range. The available...

  11. QUALITY OF COMPOSTS FROM MUNICIPAL BIODEGRADABLE WASTE

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    QUALITY OF COMPOSTS FROM MUNICIPAL BIODEGRADABLE WASTE OF DIFFERENT ORIGINS I. ZDANEVITCH AND O countries. One of the outputs of this treatment is a compost prepared from the organic matter of the waste the total MSW in the plant. Unlike in Germany or Austria, where only the compost from selective collection

  12. Biohazardous Waste Disposal Guidelines Sharps Waste Solid Lab Waste Liquid Waste Animals Human

    E-Print Network [OSTI]

    Aluwihare, Lihini

    Mixed Waste Before generating mixed waste (i.e, mixture of biohazardous and chemical or radioactive waste), call Environment, Health & Safety: (858) 534-2753. * Disinfectants other than bleach mustBiohazardous Waste Disposal Guidelines Sharps Waste Solid Lab Waste Liquid Waste Animals Human

  13. LOW ACTIVITY WASTE FEED SOLIDS CARACTERIZATION AND FILTERABILITY TESTS

    SciTech Connect (OSTI)

    McCabe, D.; Crawford, C.; Duignan, M.; Williams, M.; Burket, P.

    2014-04-03

    The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant (WTP) that is currently under construction. The baseline plan for the WTP Pretreatment facility is to treat the waste, splitting it into High Level Waste (HLW) feed and Low Activity Waste (LAW) feed. Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed onsite in the Integrated Disposal Facility (IDF). There are currently no plans to treat the waste to remove technetium in the WTP Pretreatment facility, so its disposition path is the LAW glass. Options are being explored to immobilize the LAW portion of the tank waste, i.e., the LAW feed from the WTP Pretreatment facility. Removal of {sup 99}Tc from the LAW Feed, followed by off-site disposal of the {sup 99}Tc, would eliminate a key risk contributor for the IDF Performance Assessment (PA) for supplemental waste forms, and has potential to reduce treatment and disposal costs. Washington River Protection Solutions (WRPS) is developing some conceptual flow sheets for LAW treatment and disposal that could benefit from technetium removal. One of these flowsheets will specifically examine removing {sup 99}Tc from the LAW feed stream to supplemental immobilization. The conceptual flow sheet of the {sup 99}Tc removal process includes a filter to remove insoluble solids prior to processing the stream in an ion exchange column, but the characteristics and behavior of the liquid and solid phases has not previously been investigated. This report contains results of testing of a simulant that represents the projected composition of the feed to the Supplemental LAW process. This feed composition is not identical to the aqueous tank waste fed to the Waste Treatment Plant because it has been processed through WTP Pretreatment facility and therefore contains internal changes and recycle streams that will be generated within the WTP process. Although a Supplemental LAW feed simulant has previously been prepared, this feed composition differs from that simulant because those tests examined only the fully soluble aqueous solution at room temperature, not the composition formed after evaporation, including the insoluble solids that precipitate after it cools. The conceptual flow sheet for Supplemental LAW immobilization has an option for removal of {sup 99}Tc from the feed stream, if needed. Elutable ion exchange has been selected for that process. If implemented, the stream would need filtration to remove the insoluble solids prior to processing in an ion exchange column. The characteristics, chemical speciation, physical properties, and filterability of the solids are important to judge the feasibility of the concept, and to estimate the size and cost of a facility. The insoluble solids formed during these tests were primarily natrophosphate, natroxalate, and a sodium aluminosilicate compound. At the elevated temperature and 8 M [Na+], appreciable insoluble solids (1.39 wt%) were present. Cooling to room temperature and dilution of the slurry from 8 M to 5 M [Na+] resulted in a slurry containing 0.8 wt% insoluble solids. The solids (natrophosphate, natroxalate, sodium aluminum silicate, and a hydrated sodium phosphate) were relatively stable and settled quickly. Filtration rates were in the range of those observed with iron-based simulated Hanford tank sludge simulants, e.g., 6 M [Na+] Hanford tank 241-AN-102, even though their chemical speciation is considerably different. Chemical cleaning of the crossflow filter was readily accomplished with acid. As this simulant formulation was based on an average composition of a wide range of feeds using an integrated computer model, this exact composition may never be observed. But the test conditions were selected to enable comparison to the model to enable improving its chemical prediction capability.

  14. DISSOLUTION & RESUSPENSION OF STORED RADIOACTIVE WASTE & ON SITE TRANSPORT & HANDLING FOR CONDITIONING FOR WASTE RETRIEVAL

    SciTech Connect (OSTI)

    GIBBONS, P.W.

    2001-08-13

    The four primary functions in a waste retrieval system are as follows: accessing all of the waste within the tank configuration; mobilizing all of the waste, which can have varying physical properties; removing the bulk and residual mobilized waste; and transferring the waste to storage or processing equipment. Selection of retrieval and transfer systems must include all of these functions. Limitations on any one of these areas affect the whole process. This section categorizes according to function many available retrieval and transfer processes, with positive attributes and limitations. Additional information on these systems is referenced in the annexes.

  15. Resource Conservation and Recovery Act, Part B Permit Application [for the Waste Isolation Pilot Plant (WIPP)]. Volume 2, Chapter C, Appendix C1--Chapter C, Appendix C3 (beginning), Revision 3

    SciTech Connect (OSTI)

    Not Available

    1993-03-01

    This volume contains appendices for the following: Rocky Flats Plant and Idaho National Engineering Laboratory waste process information; TRUPACT-II content codes (TRUCON); TRUPACT-II chemical list; chemical compatibility analysis for Rocky Flats Plant waste forms; chemical compatibility analysis for waste forms across all sites; TRU mixed waste characterization database; hazardous constituents of Rocky Flats Transuranic waste; summary of waste components in TRU waste sampling program at INEL; TRU waste sampling program; and waste analysis data.

  16. Characterizing cemented TRU waste for RCRA hazardous constituents

    SciTech Connect (OSTI)

    Yeamans, D.R.; Betts, S.E.; Bodenstein, S.A. [and others

    1996-06-01

    Los Alamos National Laboratory (LANL) has characterized drums of solidified transuranic (TRU) waste from four major waste streams. The data will help the State of New Mexico determine whether or not to issue a no-migration variance of the Waste Isolation Pilot Plant (WIPP) so that WIPP can receive and dispose of waste. The need to characterize TRU waste stored at LANL is driven by two additional factors: (1) the LANL RCRA Waste Analysis Plan for EPA compliant safe storage of hazardous waste; (2) the WIPP Waste Acceptance Criteria (WAC) The LANL characterization program includes headspace gas analysis, radioassay and radiography for all drums and solids sampling on a random selection of drums from each waste stream. Data are presented showing that the only identified non-metal RCRA hazardous component of the waste is methanol.

  17. A perspective of hazardous waste and mixed waste treatment technology at the Savannah River Site

    SciTech Connect (OSTI)

    England, J.L.; Venkatesh, S.; Bailey, L.L.; Langton, C.A.; Hay, M.S.; Stevens, C.B.; Carroll, S.J.

    1991-12-31

    Treatment technologies for the preparation and treatment of heavy metal mixed wastes, contaminated soils, and mixed mercury wastes are being considered at the Savannah River Site (SRS), a DOE nuclear material processing facility operated by Westinghouse Savannah River Company (WSRC). The proposed treatment technologies to be included at the Hazardous Waste/Mixed Waste Treatment Building at SRS are based on the regulatory requirements, projected waste volumes, existing technology, cost effectiveness, and project schedule. Waste sorting and size reduction are the initial step in the treatment process. After sorting/size reduction the wastes would go to the next applicable treatment module. For solid heavy metal mixed wastes the proposed treatment is macroencapsulation using a thermoplastic polymer. This process reduces the leachability of hazardous constituents from the waste and allows easy verification of the coating integrity. Stabilization and solidification in a cement matrix will treat a wide variety of wastes (i.e. soils, decontamination water). Some pretreatments may be required (i.e. Ph adjustment) before stabilization. Other pretreatments such as soil washing can reduce the amount of waste to be stabilized. Radioactive contaminated mercury waste at the SRS comes in numerous forms (i.e. process equipment, soils, and lab waste) with the required treatment of high mercury wastes being roasting/retorting and recovery. Any unrecyclable radioactive contaminated elemental mercury would be amalgamated, utilizing a batch system, before disposal.

  18. A perspective of hazardous waste and mixed waste treatment technology at the Savannah River Site

    SciTech Connect (OSTI)

    England, J.L.; Venkatesh, S.; Bailey, L.L.; Langton, C.A.; Hay, M.S.; Stevens, C.B.; Carroll, S.J.

    1991-01-01

    Treatment technologies for the preparation and treatment of heavy metal mixed wastes, contaminated soils, and mixed mercury wastes are being considered at the Savannah River Site (SRS), a DOE nuclear material processing facility operated by Westinghouse Savannah River Company (WSRC). The proposed treatment technologies to be included at the Hazardous Waste/Mixed Waste Treatment Building at SRS are based on the regulatory requirements, projected waste volumes, existing technology, cost effectiveness, and project schedule. Waste sorting and size reduction are the initial step in the treatment process. After sorting/size reduction the wastes would go to the next applicable treatment module. For solid heavy metal mixed wastes the proposed treatment is macroencapsulation using a thermoplastic polymer. This process reduces the leachability of hazardous constituents from the waste and allows easy verification of the coating integrity. Stabilization and solidification in a cement matrix will treat a wide variety of wastes (i.e. soils, decontamination water). Some pretreatments may be required (i.e. Ph adjustment) before stabilization. Other pretreatments such as soil washing can reduce the amount of waste to be stabilized. Radioactive contaminated mercury waste at the SRS comes in numerous forms (i.e. process equipment, soils, and lab waste) with the required treatment of high mercury wastes being roasting/retorting and recovery. Any unrecyclable radioactive contaminated elemental mercury would be amalgamated, utilizing a batch system, before disposal.

  19. RCRA Hazardous Waste Part A Permit Application: Instructions...

    Open Energy Info (EERE)

    Jump to: navigation, search OpenEI Reference LibraryAdd to library Form: RCRA Hazardous Waste Part A Permit Application: Instructions and Form (EPA Form 8700-23) Abstract...

  20. EIS-0082: Defense Waste Processing Facility, Savannah River Plant

    Broader source: Energy.gov [DOE]

    The Office of Defense Waste and Byproducts Management developed this EIS to provide environmental input into both the selection of an appropriate strategy for the permanent disposal of the high-level radioactive waste currently stored at the Savannah River Plant (SRP) and the subsequent decision to construct and operate a Defense Waste Processing Facility at the SRP site.

  1. WASTE TO WATTS Waste is a Resource!

    E-Print Network [OSTI]

    Columbia University

    WASTE TO WATTS Waste is a Resource! energy forum Case Studies from Estonia, Switzerland, Germany Bossart,· ABB Waste-to-Energy Plants Edmund Fleck,· ESWET Marcel van Berlo,· Afval Energie Bedrijf From Waste to Energy To Energy from Waste #12;9.00-9.30: Registration 9.30-9.40: Chairman Ella Stengler opens

  2. Nuclear waste management. Quarterly progress report, October through December 1980

    SciTech Connect (OSTI)

    Chikalla, T.D.; Powell, J.A. (comps.)

    1981-03-01

    Progress reports and summaries are presented under the following headings: high-level waste process development; alternative waste forms; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton solidification; thermal outgassing; iodine-129 fixation; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; mobility of organic complexes of radionuclides in soils; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology; high level waste form preparation; development of backfill material; development of structural engineered barriers; ONWI disposal charge analysis; spent fuel and fuel component integrity program; analysis of spent fuel policy implementation; analysis of postulated criticality events in a storage array of spent LWR fuel; asphalt emulsion sealing of uranium tailings; liner evaluation for uranium mill tailings; multilayer barriers for sealing of uranium tailings; application of long-term chemical biobarriers for uranium tailings; revegetation of inactive uranium tailing sites; verification instrument development.

  3. LITERATURE REVIEWS TO SUPPORT ION EXCHANGE TECHNOLOGY SELECTION FOR MODULAR SALT PROCESSING

    SciTech Connect (OSTI)

    King, W

    2007-11-30

    This report summarizes the results of literature reviews conducted to support the selection of a cesium removal technology for application in a small column ion exchange (SCIX) unit supported within a high level waste tank. SCIX is being considered as a technology for the treatment of radioactive salt solutions in order to accelerate closure of waste tanks at the Savannah River Site (SRS) as part of the Modular Salt Processing (MSP) technology development program. Two ion exchange materials, spherical Resorcinol-Formaldehyde (RF) and engineered Crystalline Silicotitanate (CST), are being considered for use within the SCIX unit. Both ion exchange materials have been studied extensively and are known to have high affinities for cesium ions in caustic tank waste supernates. RF is an elutable organic resin and CST is a non-elutable inorganic material. Waste treatment processes developed for the two technologies will differ with regard to solutions processed, secondary waste streams generated, optimum column size, and waste throughput. Pertinent references, anticipated processing sequences for utilization in waste treatment, gaps in the available data, and technical comparisons will be provided for the two ion exchange materials to assist in technology selection for SCIX. The engineered, granular form of CST (UOP IE-911) was the baseline ion exchange material used for the initial development and design of the SRS SCIX process (McCabe, 2005). To date, in-tank SCIX has not been implemented for treatment of radioactive waste solutions at SRS. Since initial development and consideration of SCIX for SRS waste treatment an alternative technology has been developed as part of the River Protection Project Waste Treatment Plant (RPP-WTP) Research and Technology program (Thorson, 2006). Spherical RF resin is the baseline media for cesium removal in the RPP-WTP, which was designed for the treatment of radioactive waste supernates and is currently under construction in Hanford, WA. Application of RF for cesium removal in the Hanford WTP does not involve in-riser columns but does utilize the resin in large scale column configurations in a waste treatment facility. The basic conceptual design for SCIX involves the dissolution of saltcake in SRS Tanks 1-3 to give approximately 6 M sodium solutions and the treatment of these solutions for cesium removal using one or two columns supported within a high level waste tank. Prior to ion exchange treatment, the solutions will be filtered for removal of entrained solids. In addition to Tanks 1-3, solutions in two other tanks (37 and 41) will require treatment for cesium removal in the SCIX unit. The previous SCIX design (McCabe, 2005) utilized CST for cesium removal with downflow supernate processing and included a CST grinder following cesium loading. Grinding of CST was necessary to make the cesium-loaded material suitable for vitrification in the SRS Defense Waste Processing Facility (DWPF). Because RF resin is elutable (and reusable) and processing requires conversion between sodium and hydrogen forms using caustic and acidic solutions more liquid processing steps are involved. The WTP baseline process involves a series of caustic and acidic solutions (downflow processing) with water washes between pH transitions across neutral. In addition, due to resin swelling during conversion from hydrogen to sodium form an upflow caustic regeneration step is required. Presumably, one of these basic processes (or some variation) will be utilized for MSP for the appropriate ion exchange technology selected. CST processing involves two primary waste products: loaded CST and decontaminated salt solution (DSS). RF processing involves three primary waste products: spent RF resin, DSS, and acidic cesium eluate, although the resin is reusable and typically does not require replacement until completion of multiple treatment cycles. CST processing requires grinding of the ion exchange media, handling of solids with high cesium loading, and handling of liquid wash and conditioning solutions. RF processing requires h

  4. HAZARDOUS WASTE [Written Program

    E-Print Network [OSTI]

    Pawlowski, Wojtek

    HAZARDOUS WASTE MANUAL [Written Program] Cornell University [10/7/13 #12;Hazardous Waste Program................................................... 8 3.0 MINIMIZING HAZARDOUS WASTE GENERATION.........................................................10 4.0 HAZARDOUS WASTE GENERATOR REQUIREMENTS.....................................................10

  5. Waste Management

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power AdministrationRobust,Field-effectWorking WithTelecentricN A 035(92/02) nergFeet)DepartmentWasteWaste

  6. DM100 AND DM1200 MELTER TESTING WITH HIGH WASTE LOADING GLASS FORMULATIONS FOR HANFORD HIGH-ALUMINUM HLW STREAMS

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT WK; PEGG IL; JOSEPH I

    2009-12-30

    This Test Plan describes work to support the development and testing of high waste loading glass formulations that achieve high glass melting rates for Hanford high aluminum high level waste (HLW). In particular, the present testing is designed to evaluate the effect of using low activity waste (LAW) waste streams as a source of sodium in place ofchemical additives, sugar or cellulose as a reductant, boehmite as an aluminum source, and further enhancements to waste processing rate while meeting all processing and product quality requirements. The work will include preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM 100) tests designed to examine the effects of enhanced glass formulations, glass processing temperature, incorporation of the LAW waste stream as a sodium source, type of organic reductant, and feed solids content on waste processing rate and product quality. Also included is a confirmatory test on the HLW Pilot Melter (DM1200) with a composition selected from those tested on the DM100. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy's (DOE's) Office of River Protection (ORP) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same waste composition. This Test Plan is prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is about 12,500. This estimate is based upon the inventory ofthe tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP contract requirements. The WTP's overall mission will require the immobilization oftank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in waste-loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  7. Electronic waste disassembly with industrial waste heat

    E-Print Network [OSTI]

    2013-01-01

    and for e?ective use of industrial exhaust heat is describedto scale up the process to industrial production levels.Waste Disassembly with Industrial Waste Heat Mengjun

  8. Data summary of municipal solid waste management alternatives. Volume 3, Appendix A: Mass burn technologies

    SciTech Connect (OSTI)

    1992-10-01

    This appendix on Mass Burn Technologies is the first in a series designed to identify, describe and assess the suitability of several currently or potentially available generic technologies for the management of municipal solid waste (MSW). These appendices, which cover eight core thermoconversion, bioconversion and recycling technologies, reflect public domain information gathered from many sources. Representative sources include: professional journal articles, conference proceedings, selected municipality solid waste management plans and subscription technology data bases. The information presented is intended to serve as background information that will facilitate the preparation of the technoeconomic and life cycle mass, energy and environmental analyses that are being developed for each of the technologies. Mass burn has been and continues to be the predominant technology in Europe for the management of MSW. In the United States, the majority of the existing waste-to-energy projects utilize this technology and nearly 90 percent of all currently planned facilities have selected mass burn systems. Mass burning generally refers to the direct feeding and combustion of municipal solid waste in a furnace without any significant waste preprocessing. The only materials typically removed from the waste stream prior to combustion are large bulky objects and potentially hazardous or undesirable wastes. The technology has evolved over the last 100 or so years from simple incineration to the most highly developed and commercially proven process available for both reducing the volume of MSW and for recovering energy in the forms of steam and electricity. In general, mass burn plants are considered to operate reliably with high availability.

  9. An expert system framework for nondestructive waste assay

    SciTech Connect (OSTI)

    Becker, G.K.

    1996-10-01

    Management and disposition of transuranic (RU) waste forms necessitates determining entrained RU and associated radioactive material quantities as per National RU Waste Characterization Program requirements. Technical justification and demonstration of a given NDA method used to determine RU mass and uncertainty in accordance with program quality assurance is difficult for many waste forms. Difficulties are typically founded in waste NDA methods that employ standards compensation and/or employment of simplifying assumptions on waste form configurations. Capability to determine and justify RU mass and mass uncertainty can be enhanced through integration of waste container data/information using expert system and empirical data-driven techniques with conventional data acquisition and analysis. Presented is a preliminary expert system framework that integrates the waste form data base, alogrithmic techniques, statistical analyses, expert domain knowledge bases, and empirical artificial intelligence modules into a cohesive system. The framework design and bases in addition to module development activities are discussed.

  10. Small businesses selected for nuclear waste services

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power AdministrationRobust,Field-effect Photovoltaics -7541 *Impact NeutronSmall Business- NewsSmallMore than

  11. Radioactive waste processing apparatus

    DOE Patents [OSTI]

    Nelson, Robert E. (Lombard, IL); Ziegler, Anton A. (Darien, IL); Serino, David F. (Maplewood, MN); Basnar, Paul J. (Western Springs, IL)

    1987-01-01

    Apparatus for use in processing radioactive waste materials for shipment and storage in solid form in a container is disclosed. The container includes a top, and an opening in the top which is smaller than the outer circumference of the container. The apparatus includes an enclosure into which the container is placed, solution feed apparatus for adding a solution containing radioactive waste materials into the container through the container opening, and at least one rotatable blade for blending the solution with a fixing agent such as cement or the like as the solution is added into the container. The blade is constructed so that it can pass through the opening in the top of the container. The rotational axis of the blade is displaced from the center of the blade so that after the blade passes through the opening, the blade and container can be adjusted so that one edge of the blade is adjacent the cylindrical wall of the container, to insure thorough mixing. When the blade is inside the container, a substantially sealed chamber is formed to contain vapors created by the chemical action of the waste solution and fixant, and vapors emanating through the opening in the container.

  12. Biohazardous Waste Disposal Guidelines Sharps Waste Solid Lab Waste Liquid Waste Animals Human

    E-Print Network [OSTI]

    Aluwihare, Lihini

    of biohazardous and chemical or radioactive waste), call Environment, Health & Safety: (858) 534Biohazardous Waste Disposal Guidelines Sharps Waste Solid Lab Waste Liquid Waste Animals Human Pathological Waste Description Biohazard symbol Address: UCSD 200 West Arbor Dr. San Diego, CA 92103 (858

  13. Nondestructive assay of boxed radioactive waste

    SciTech Connect (OSTI)

    Gilles, W.P.; Roberts, R.J.; Jasen, W.G.

    1992-12-01

    This paper describes the problems related to the nondestructive assay (NDA) of boxed radioactive waste at the Hanford Site and how Westinghouse Hanford company (WHC) is solving the problems. The waste form and radionuclide content are described. The characteristics of the combined neutron and gamma-based measurement system are described.

  14. Vitrification technology for Hanford Site tank waste

    SciTech Connect (OSTI)

    Weber, E.T.; Calmus, R.B.; Wilson, C.N.

    1995-04-01

    The US Department of Energy`s (DOE) Hanford Site has an inventory of 217,000 m{sup 3} of nuclear waste stored in 177 underground tanks. The DOE, the US Environmental Protection Agency, and the Washington State Department of Ecology have agreed that most of the Hanford Site tank waste will be immobilized by vitrification before final disposal. This will be accomplished by separating the tank waste into high- and low-level fractions. Capabilities for high-capacity vitrification are being assessed and developed for each waste fraction. This paper provides an overview of the program for selecting preferred high-level waste melter and feed processing technologies for use in Hanford Site tank waste processing.

  15. Polymer filtration: An emerging technology for selective metals recovery

    SciTech Connect (OSTI)

    Smith, B.F.; Robison, T.W.; Cournoyer, M.E.

    1995-12-31

    A new technology is under development to selectively recover regulated metal ions from electroplating rinse waters. The electroplating metal ions are recovered in a concentrated form with the appropriate counter ions ready for return to the original electroplating bath. The technology is based on the use of specially designed water-soluble polymers that selectively bind with the metal ions in the rinse bath. The polymers have such a large molecular weight that they can be physically separated using available ultrafiltration technology. The advantages of this technology are high metal selectivity with no sludge formation, rapid processing, low energy, low capital costs, and small size. We have tested and demonstrated the recovery of zinc and nickel (a new alloy electroplating bath designed to replace cadmium) from rinse waters. The metal-ion concentrate was returned to the original electroplating bath. Applications of this technology include waste treatment for textile, paint and dye production, chemical manufacturing, and nuclear reactor and reprocessing operations.

  16. ORS 466 - Storage, Treatment, and Disposal of Hazardous Waste...

    Open Energy Info (EERE)

    Apps Datasets Community Login | Sign Up Search Page Edit with form History ORS 466 - Storage, Treatment, and Disposal of Hazardous Waste and Materials Jump to: navigation, search...

  17. Radiological, physical, and chemical characterization of transuranic wastes stored at the Idaho National Engineering Laboratory

    SciTech Connect (OSTI)

    Apel, M.L.; Becker, G.K.; Ragan, Z.K.; Frasure, J.; Raivo, B.D.; Gale, L.G.; Pace, D.P.

    1994-03-01

    This document provides radiological, physical and chemical characterization data for transuranic radioactive wastes and transuranic radioactive and hazardous (i.e., mixed) wastes stored at the Idaho National Engineering Laboratory and considered for treatment under the Private Sector Participation Initiative Program (PSPI). Waste characterization data are provided in the form of INEL Waste Profile Sheets. These documents provide, for each content code, information on waste identification, waste description, waste storage configuration, physical/chemical waste composition, radionuclide and associated alpha activity waste characterization data, and hazardous constituents present in the waste. Information is provided for 139 waste streams which represent an estimated total volume of 39,380{sup 3} corresponding to a total mass of approximately 19,000,000 kg. In addition, considerable information concerning alpha, beta, gamma, and neutron source term data specific to Rocky Flats Plant generated waste forms stored at the INEL are provided to assist in facility design specification.

  18. Risk Assessment supporting the decision on the initial selection of supplemental ILAW technologies

    SciTech Connect (OSTI)

    MANN, F. M.

    2003-09-29

    A risk assessment on the long-term environmental impact of various potential waste forms was conducted at the request of the Hanford Site's Mission Acceleration Initiative Team. These potential waste forms (bulk vitrification, cast stone, and steam reformer) may treat some of the low-activity waste currently planned to be treated at the Waste Treatment Plant.

  19. Tank farms compacted low-level waste

    SciTech Connect (OSTI)

    Hetzer, D.C.

    1997-08-01

    This report describes the process of Low-Level Waste (LLW) volume reduction by compaction. Also included is the data used for characterization of LLW destined for compaction. Scaling factors (ratios) are formed based on data contained in this report.

  20. Tank farms compacted low level waste

    SciTech Connect (OSTI)

    Waters, M.S., Westinghouse Hanford

    1996-07-01

    This report describes the process of Low Level Waste (LLW) volume reduction by compaction. Also included is the data used for characterization of LLW destined for compaction. Scaling factors (ratios) are formed based on data contained in this report.

  1. WASTE DISPOSAL WORKSHOPS: ANTHRAX CONTAMINATED WASTE

    E-Print Network [OSTI]

    large amounts of waste that must be managed as part of both immediate recovery and long-term recovery management plans that can address contaminated waste through the entire life cycle of the waste. Through Demonstration LLNL Lawrence Livermore National Laboratory MSW Municipal Solid Waste OSHA Occupational Safety

  2. Second-order model selection in mixture experiments

    SciTech Connect (OSTI)

    Redgate, P.E.; Piepel, G.F.; Hrma, P.R.

    1992-07-01

    Full second-order models for q-component mixture experiments contain q(q+l)/2 terms, which increases rapidly as q increases. Fitting full second-order models for larger q may involve problems with ill-conditioning and overfitting. These problems can be remedied by transforming the mixture components and/or fitting reduced forms of the full second-order mixture model. Various component transformation and model reduction approaches are discussed. Data from a 10-component nuclear waste glass study are used to illustrate ill-conditioning and overfitting problems that can be encountered when fitting a full second-order mixture model. Component transformation, model term selection, and model evaluation/validation techniques are discussed and illustrated for the waste glass example.

  3. Process for producing zeolite adsorbent and process for treating radioactive liquid waste with the zeolite adsorbent

    SciTech Connect (OSTI)

    Motojima, K.; Kawamura, F.

    1984-05-15

    Zeolite is contacted with an aqueous solution containing at least one of copper, nickel, cobalt, manganese and zinc salts, preferably copper and nickel salts, particularly preferably copper salt, in such a form as sulfate, nitrate, or chloride, thereby adsorbing the metal on the zeolite in its pores by ion exchange, then the zeolite is treated with a water-soluble ferrocyanide compound, for example, potassium ferrocyanide, thereby forming metal ferrocyanide on the zeolite in its pores. Then, the zeolite is subjected to ageing treatment, thereby producing a zeolite adsorbent impregnated with metal ferrocyanide in the pores of zeolite. The adsorbent can selectively recover cesium with a high percent cesium removal from a radioactive liquid waste containing at least radioactive cesium, for example, a radioactive liquid waste containing cesium and such coexisting ions as sodium, magnesium, calcium and carbonate ions at the same time at a high concentration. The zeolite adsorbent has a stable adsorbability for a prolonged time.

  4. Sodium Recycle Economics for Waste Treatment Plant Operations

    SciTech Connect (OSTI)

    Sevigny, Gary J.; Poloski, Adam P.; Fountain, Matthew S.

    2008-08-31

    Sodium recycle at the Hanford Waste Treatment Plant (WTP) would reduce the number of glass canisters produced, and has the potential to significantly reduce the cost to the U.S. Department of Energy (DOE) of treating the tank wastes by hundreds of millions of dollars. The sodium, added in the form of sodium hydroxide, was originally added to minimize corrosion of carbon-steel storage tanks from acidic reprocessing wastes. In the baseline Hanford treatment process, sodium hydroxide is required to leach gibbsite and boehmite from the high level waste (HLW) sludge. In turn, this reduces the amount of HLW glass produced. Currently, a significant amount of additional sodium hydroxide will be added to the process to maintain aluminate solubility at ambient temperatures during ion exchange of cesium. The vitrification of radioactive waste is limited by sodium content, and this additional sodium mass will increase low-activity waste-glass mass. An electrochemical salt-splitting process, based on sodium-ion selective ceramic membranes, is being developed to recover and recycle sodium hydroxide from high-salt radioactive tank wastes in DOE’s complex. The ceramic membranes are from a family of materials known as sodium (Na)—super-ionic conductors (NaSICON)—and the diffusion of sodium ions (Na+) is allowed, while blocking other positively charged ions. A cost/benefit evaluation was based on a strategy that involves a separate caustic-recycle facility based on the NaSICON technology, which would be located adjacent to the WTP facility. A Monte Carlo approach was taken, and several thousand scenarios were analyzed to determine likely economic results. The cost/benefit evaluation indicates that 10,000–50,000 metric tons (MT) of sodium could be recycled, and would allow for the reduction of glass production by 60,000–300,000 MT. The cost of the facility construction and operation was scaled to the low-activity waste (LAW) vitrification facility, showing cost would be roughly $150 million to $400 million for construction and $10 million to $40 million per year for operations. Depending on the level of aluminate supersaturation allowed in the storage tanks in the LAW Pretreatment Facility, these values indicate a return on investment of up to 25% to 60%.

  5. Purchasing Services SUPPLIER SELECTION JUSTIFICATION

    E-Print Network [OSTI]

    Doering, Tamara

    Purchasing Services SUPPLIER SELECTION JUSTIFICATION Instructions: This form must be completed,000. The elements to be satisfied are: 1) cost basis (bids or quote); 2) justification/rationale for the supplier selection. FIS Document #:________________ Supplier Name: ______________________________ Product Description

  6. Remaining Sites Verification Package for the 100-B-21:2 Subsite (100-B/C Discovery Pipeline DS-100BC-002), Waste Site Reclassification Form 2008-003

    SciTech Connect (OSTI)

    J. M. Capron

    2008-06-16

    The 100-B-21:2 waste site consists of the immediate area of the DS-100BC-02 pipeline. In accordance with this evaluation, the confirmatory and verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  7. NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA, JUNE 2006

    SciTech Connect (OSTI)

    U.S. DEPARTMENT OF ENERGY, NATIONAL NUCLEAR SECURITY ADMINISTRATION NEVADA SITE OFFICE

    2006-06-01

    This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive (LLW) and mixed waste (MW) for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex (RWMC) for storage or disposal.

  8. High Level Waste Disposal System Optimization

    SciTech Connect (OSTI)

    Dirk Gombert; M. Connolly; J. Roach; W. Holtzscheiter

    2005-02-01

    The high level waste (HLW) disposal system consists of the Yucca Mountain Facility (YMF) and waste product (e.g. glass) generation facilities. Responsibility for management is shared between the U. S. Department of Energy (DOE) Offices of Civilian Radioactive Waste Management (DOE-RW) and Environmental Management (DOE-EM). The DOE-RW license application and the Waste Acceptance System Requirements Document (WASRD), as well as the DOE-EM Waste Acceptance Product Specification for Vitrified High Level Waste Forms (WAPS) govern the overall performance of the system. This basis for HLW disposal should be reassessed to consider waste form and process technology research and development (R&D), which have been conducted by DOE-EM, international agencies (i.e. ANSTO, CEA), and the private sector; as well as the technical bases for including additional waste forms in the final license application. This will yield a more optimized HLW disposal system to accelerate HLW disposition, more efficient utilization of the YMF, and overall system cost reduction.

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

    SciTech Connect (OSTI)

    Fox, K.; Brinkman, K.

    2011-09-22

    The Savannah River National Laboratory (SRNL) is developing crystalline ceramic waste forms to incorporate CS/LN/TM high Mo waste streams consisting of perovskite, hollandite, pyrochlore, zirconolite, and powellite phase assemblages. Simple raw materials, including Al{sub 2}O{sub 3}, CaO, and TiO{sub 2} were combined with simulated waste components to produce multiphase crystalline ceramics. Fiscal Year 2011 (FY11) activities included (i) expanding the compositional range by varying waste loading and fabrication of compositions rich in TiO{sub 2}, (ii) exploring the processing parameters of ceramics produced by the melt and crystallize process, (iii) synthesis and characterization of select individual phases of powellite and hollandite that are the target hosts for radionuclides of Mo, Cs, and Rb, and (iv) evaluating the durability and radiation stability of single and multi-phase ceramic waste forms. Two fabrication methods, including melting and crystallizing, and pressing and sintering, were used with the intent of studying phase evolution under various sintering conditions. An analysis of the XRD and SEM/EDS results indicates that the targeted crystalline phases of the FY11 compositions consisting of pyrochlore, perovskite, hollandite, zirconolite, and powellite were formed by both press and sinter and melt and crystallize processing methods. An evaluation of crystalline phase formation versus melt processing conditions revealed that hollandite, perovskite, zirconolite, and residual TiO{sub 2} phases formed regardless of cooling rate, demonstrating the robust nature of this process for crystalline phase development. The multiphase ceramic composition CSLNTM-06 demonstrated good resistance to proton beam irradiation. Electron irradiation studies on the single phase CaMoO{sub 4} (a component of the multiphase waste form) suggested that this material exhibits stability to 1000 years at anticipated self-irradiation doses (2 x 10{sup 10}-2 x 10{sup 11} Gy), but that its stability may be rate dependent, therefore limiting the activity of the waste for which it can be employed. Overall, these preliminary results indicate good radiation damage tolerance for the crystalline ceramic materials. The PCT results showed that, for all of the waste forms tested, the normalized release values for most of the elements measured, including all of the lanthanides and noble metals, were either very small or below the instrument detection limits. Elevated normalized release values were measured only for Cs, Mo, and Rb. It is difficult to draw further conclusions from these data until a benchmark material is developed for the PCT with this type of waste form. Calcined, simulated CS/LN/TM High Mo waste without additives had relatively low normalized release values for Cs, Mo, and Rb. A review of the chemical composition data for this sample showed that these elements were well retained after the calcination. Therefore, it will be useful to further characterize the calcined material to determine what form these elements are in after calcining. This, along with single phase studies on Cs containing crystal structures such as hollandite, should provide insight into the most ideal phases to incorporate these elements to produce a durable waste form.

  10. Radioactive waste processing apparatus

    DOE Patents [OSTI]

    Nelson, R.E.; Ziegler, A.A.; Serino, D.F.; Basnar, P.J.

    1985-08-30

    Apparatus for use in processing radioactive waste materials for shipment and storage in solid form in a container is disclosed. The container includes a top, and an opening in the top which is smaller than the outer circumference of the container. The apparatus includes an enclosure into which the container is placed, solution feed apparatus for adding a solution containing radioactive waste materials into the container through the container opening, and at least one rotatable blade for blending the solution with a fixing agent such as cement or the like as the solution is added into the container. The blade is constructed so that it can pass through the opening in the top of the container. The rotational axis of the blade is displaced from the center of the blade so that after the blade passes through the opening, the blade and container can be adjusted so that one edge of the blade is adjacent the cylindrical wall of the container, to insure thorough mixing. When the blade is inside the container, a substantially sealed chamber is formed to contain vapors created by the chemical action of the waste solution and fixant, and vapors emanating through the opening in the container. The chamber may be formed by placing a removable extension over the top of the container. The extension communicates with the apparatus so that such vapors are contained within the container, extension and solution feed apparatus. A portion of the chamber includes coolant which condenses the vapors. The resulting condensate is returned to the container by the force of gravity.

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

    SciTech Connect (OSTI)

    Hamel, W. F. [Office of River Protection, U.S. Department of Energy, 2400 Stevens Drive, Richland, WA 99354 (United States); Gerdes, K. [U.S. Department of Energy, 19901 Germantown Road, Germantown, MD 20874 (United States); Holton, L. K. [Pacific Northwest National Laboratory, PO Box 999, Richland, WA 99352 (United States); Pegg, I.L. [Vitreous State Laboratory, The Catholic University of America, 620 Michigan Avenue NE, Washington, DC 20064 (United States); Bowan, B.W. [Duratek, Inc., 10100 Old Columbia Road, Columbia, Maryland 21046 (United States)

    2006-07-01

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

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

    SciTech Connect (OSTI)

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

    1980-09-01

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

  13. Leaching Characteristics of Hanford Ferrocyanide Wastes

    SciTech Connect (OSTI)

    Edwards, Matthew K.; Fiskum, Sandra K.; Peterson, Reid A.; Shimskey, Rick W.

    2009-12-21

    A series of leach tests were performed on actual Hanford Site tank wastes in support of the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The samples were targeted composite slurries of high-level tank waste materials representing major complex, radioactive, tank waste mixtures at the Hanford Site. Using a filtration/leaching apparatus, sample solids were concentrated, caustic leached, and washed under conditions representative of those planned for the Pretreatment Facility in the WTP. Caustic leaching was performed to assess the mobilization of aluminum (as gibbsite, Al[OH]3, and boehmite AlO[OH]), phosphates [PO43-], chromium [Cr3+] and, to a lesser extent, oxalates [C2O42-]). Ferrocyanide waste released the solid phase 137Cs during caustic leaching; this was antithetical to the other Hanford waste types studied. Previous testing on ferrocyanide tank waste focused on the aging of the ferrocyanide salt complex and its thermal compatibilities with nitrites and nitrates. Few studies, however, examined cesium mobilization in the waste. Careful consideration should be given to the pretreatment of ferrocyanide wastes in light of this new observed behavior, given the fact that previous testing on simulants indicates a vastly different cesium mobility in this waste form. The discourse of this work will address the overall ferrocyanide leaching characteristics as well as the behavior of the 137Cs during leaching.

  14. Automated Sorting of Transuranic Waste

    SciTech Connect (OSTI)

    Shurtliff, Rodney Marvin

    2001-03-01

    The HANDSS-55 Transuranic Waste Sorting Module is designed to sort out items found in 55-gallon drums of waste as determined by an operator. Innovative imaging techniques coupled with fast linear motor-based motion systems and a flexible end-effector system allow the operator to remove items from the waste stream by a touch of the finger. When all desired items are removed from the waste stream, the remaining objects are automatically moved to a repackaging port for removal from the glovebox/cell. The Transuranic Waste Sorting Module consists of 1) a high accuracy XYZ Stereo Measurement and Imaging system, 2) a vibrating/tilting sorting table, 3) an XY Deployment System, 4) a ZR Deployment System, 5) several user-selectable end-effectors, 6) a waste bag opening system, 7) control and instrumentation, 8) a noncompliant waste load-out area, and 9) a Human/Machine Interface (HMI). The system is modular in design to accommodate database management tools, additional load-out ports, and other enhancements. Manually sorting the contents of a 55-gallon drum takes about one day per drum. The HANDSS-55 Waste Sorting Module is designed to significantly increase the throughput of this sorting process by automating those functions that are strenuous and tiresome for an operator to perform. The Waste Sorting Module uses the inherent ability of an operator to identify the items that need to be segregated from the waste stream and then, under computer control, picks that item out of the waste and deposits it in the appropriate location. The operator identifies the object by locating the visual image on a large color display and touches the image on the display with his finger. The computer then determines the location of the object, and performing a highspeed image analysis determines its size and orientation, so that a robotic gripper can be deployed to pick it up. Following operator verification by voice or function key, the object is deposited into a specified location.

  15. Hanford Site annual dangerous waste report: Volume 1, Part 1, Generator dangerous waste report, dangerous waste

    SciTech Connect (OSTI)

    NONE

    1994-12-31

    This report contains information on hazardous wastes at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, waste number, weight, and waste designation.

  16. Remaining Sites Verification Package for the 100-F-26:10, 1607-F3 Sanitary Sewer Pipelines (182-F, 183-F, and 151-F Sanitary Sewer Lines), Waste Site Reclassification Form 2007-028

    SciTech Connect (OSTI)

    L. M. Dittmer

    2007-12-03

    The 100-F-26:10 waste site includes sanitary sewer lines that serviced the former 182-F, 183-F, and 151-F Buildings. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  17. Remaining Sites Verification Package for the 100-F-36, 108-F Biological Laboratory, and for the 116-F-15, 108-F Radiation Crib, Waste Site Reclassification Form 2007-002

    SciTech Connect (OSTI)

    L. M. Dittmer

    2007-05-24

    The 100-F-36 waste site is the location of the former 108-F Biological Laboratory. The building was closed in 1973, decontaminated, decommissioned, and eventually demolished in 1999. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  18. Preliminary waste acceptance criteria for the ICPP spent fuel and waste management technology development program

    SciTech Connect (OSTI)

    Taylor, L.L.; Shikashio, R.

    1993-09-01

    The purpose of this document is to identify requirements to be met by the Producer/Shipper of Spent Nuclear Fuel/High-LeveL Waste SNF/HLW in order for DOE to be able to accept the packaged materials. This includes defining both standard and nonstandard waste forms.

  19. Economic evaluation of volume reduction for Defense transuranic waste

    SciTech Connect (OSTI)

    Brown, C.M.

    1981-07-01

    This study evaluates the economics of volume reduction of retrievably stored and newly generated DOE transuranic waste by comparing the costs of reduction of the waste with the savings possible in transportation and disposal of the waste. The report develops a general approach to the comparison of TRU waste volume reduction costs and cost savings, establishes an initial set of cost data, and develops conclusions to support selecting technologies and facilities for the disposal of DOE transuranic waste. Section I outlines the analysis which considers seven types of volume reduction from incineration and compaction of combustibles to compaction, size reduction, shredding, melting, and decontamination of metals. The study considers the volume reduction of contact-handled newly generated, and retrievably stored DOE transuranic waste. Section II of this report describes the analytical approach, assumptions, and flow of waste material through sites. Section III presents the waste inventories, disposal, and transportation savings with volume reduction and the volume reduction techniques and savings.

  20. Sorbent Testing for the Solidification of Unidentified Rocky Flats Laboratory Waste Stored at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Bickford, J.; Kimmitt, R.

    2007-07-01

    At the request of the U.S. Department of Energy (DOE), MSE Technology Applications, Inc. (MSE) evaluated various commercially available sorbents to solidify unidentified laboratory liquids from Rocky Flats that are stored at the Idaho National Laboratory (INL). The liquids are a collection of laboratory wastes that were generated from various experiments and routine analytical laboratory activities carried out at Rocky Flats. The liquids are in bottles discovered inside of buried waste drums being exhumed from the subsurface disposal area at the Radioactive Waste Management Complex (RWMC) by the contractor, CH2M Hill Washington International (CWI). Free liquids are unacceptable at the Waste Isolation Pilot Plant (WIPP), and some of these liquids cannot be returned to the retrieval pit. Stabilization of the liquids into a solid mass will allow these materials to be sent to an appropriate disposal location. The selected sorbent or sorbent combinations should produce a stabilized mass that is capable of withstanding conditions similar to those experienced during storage, shipping, and burial. The final wasteform should release less than 1% liquid by volume per the WIPP Waste Acceptance Criteria (WAC). The absence or presence of free liquid in the solidified waste-forms was detected when tested by SW-846, Method 9095B, Paint Filter Free Liquids, and the amount of liquid released from the wasteform was determined by SW-846, Method 9096, Liquid Release Test. Reactivity testing was also conducted on the solidified laboratory liquids. (authors)

  1. Superconducting open-gradient magnetic separation for the pretreatment of radioactive or mixed waste vitrification feeds. 1997 annual progress report

    SciTech Connect (OSTI)

    Doctor, R.; Nunez, L. [Argonne National Lab., IL (US); Cicero-Herman, C.A. [Westinghouse Savannah River Co., Aiken, SC (US). Savannah River Technology Center; Ritter, J.A. [Univ. of South Carolina, Columbia, SC (US). Chemical Engineering Dept.; Landsberger, S. [Univ. of Texas, Austin, TX (US). Nuclear Engineering Dept.

    1997-01-01

    'Vitrification has been selected as a final waste form technology in the US for long-term storage of high-level radioactive wastes (HLW). However, a foreseeable problem during vitrification in some waste feed streams lies in the presence of elements (e.g., transition metals) in the HLW that may cause instabilities in the final glass product. The formation of spinel compounds, such as Fe{sub 3}O{sub 4} and FeCrO{sub 4}, results in glass phase separation and reduces vitrifier lifetime, and durability of the final waste form. A superconducting open gradient magnetic separation (OGMS) system maybe suitable for the removal of the deleterious transition elements (e.g. Fe, Co, and Ni) and other elements (lanthanides) from vitrification feed streams due to their ferromagnetic or paramagnetic nature. The OGMS systems are designed to deflect and collect paramagnetic minerals as they interact with a magnetic field gradient. This system has the potential to reduce the volume of HLW for vitrification and ensure a stable product. In order to design efficient OGMS and High gradient magnetic separation (HGMS) processes, a fundamental understanding of the physical and chemical properties of the waste feed streams is required. Using HLW simulant and radioactive fly ash and sludge samples from the Savannah River Technology Center, Rocky Flats site, and the Hanford reservation, several techniques were used to characterize and predict the separation capability for a superconducting OGMS system.'

  2. Plasma vitrification of waste materials

    DOE Patents [OSTI]

    McLaughlin, D.F.; Dighe, S.V.; Gass, W.R.

    1997-06-10

    This invention provides a process wherein hazardous or radioactive wastes in the form of liquids, slurries, or finely divided solids are mixed with finely divided glassformers (silica, alumina, soda, etc.) and injected directly into the plume of a non-transferred arc plasma torch. The extremely high temperatures and heat transfer rates makes it possible to convert the waste-glassformer mixture into a fully vitrified molten glass product in a matter of milliseconds. The molten product may then be collected in a crucible for casting into final wasteform geometry, quenching in water, or further holding time to improve homogeneity and eliminate bubbles. 4 figs.

  3. Plasma vitrification of waste materials

    DOE Patents [OSTI]

    McLaughlin, David F. (Oakmont, PA); Dighe, Shyam V. (North Huntingdon, PA); Gass, William R. (Plum Boro, PA)

    1997-01-01

    This invention provides a process wherein hazardous or radioactive wastes in the form of liquids, slurries, or finely divided solids are mixed with finely divided glassformers (silica, alumina, soda, etc.) and injected directly into the plume of a non-transferred arc plasma torch. The extremely high temperatures and heat transfer rates makes it possible to convert the waste-glassformer mixture into a fully vitrified molten glass product in a matter of milliseconds. The molten product may then be collected in a crucible for casting into final wasteform geometry, quenching in water, or further holding time to improve homogeneity and eliminate bubbles.

  4. Systems engineering identification and control of mixed waste technology development

    SciTech Connect (OSTI)

    Beitel, G.A.

    1997-08-01

    The Department of Energy (DOE) established the Mixed Waste Characterization, Treatment, and Disposal Focus Area (MWFA) to develop technologies required to meet the Department`s commitments for treatment of mixed low-level and transuranic wastes. Waste treatment includes all necessary steps from generation through disposal. Systems engineering was employed to reduce programmatic risk, that is, risk of failure to meet technical commitments within cost and schedule. Customer needs (technology deficiencies) are identified from Site Treatment Plans, Consent Orders, ten year plans, Site Technical Coordinating Groups, Stakeholders, and Site Visits. The Technical Baseline, a prioritized list of technology deficiencies, forms the basis for determining which technology development activities will be supported by the MWFA. Technology Development Requirements Documents are prepared for each technology selected for development. After technologies have been successfully developed and demonstrated, they are documented in a Technology Performance Report. The Technology Performance Reports are available to any of the customers or potential users of the technology, thus closing the loop between problem identification and product development. This systematic approach to technology development and its effectiveness after 3 years is discussed in this paper.

  5. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    SciTech Connect (OSTI)

    T.L. Lotz

    1997-02-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercial spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.

  6. Waste remediation

    DOE Patents [OSTI]

    Halas, Nancy J.; Nordlander, Peter; Neumann, Oara

    2015-12-29

    A system including a steam generation system and a chamber. The steam generation system includes a complex and the steam generation system is configured to receive water, concentrate electromagnetic (EM) radiation received from an EM radiation source, apply the EM radiation to the complex, where the complex absorbs the EM radiation to generate heat, and transform, using the heat generated by the complex, the water to steam. The chamber is configured to receive the steam and an object, wherein the object is of medical waste, medical equipment, fabric, and fecal matter.

  7. Method for forming metal contacts

    DOE Patents [OSTI]

    Reddington, Erik; Sutter, Thomas C; Bu, Lujia; Cannon, Alexandra; Habas, Susan E; Curtis, Calvin J; Miedaner, Alexander; Ginley, David S; Van Hest, Marinus Franciscus Antonius Maria

    2013-09-17

    Methods of forming metal contacts with metal inks in the manufacture of photovoltaic devices are disclosed. The metal inks are selectively deposited on semiconductor coatings by inkjet and aerosol apparatus. The composite is heated to selective temperatures where the metal inks burn through the coating to form an electrical contact with the semiconductor. Metal layers are then deposited on the electrical contacts by light induced or light assisted plating.

  8. Finding of no significant impact for the interim action for cleanup of Pit 9 at the Radioactive Waste Management Complex, Idaho National Engineering Laboratory

    SciTech Connect (OSTI)

    Not Available

    1993-10-01

    The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0854, for an interim action under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). The proposed action would be conducted at Pit 9, Operable Unit 7--10, located at the Subsurface Disposal Area (SDA) of the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering Laboratory (INEL). The proposed action consists of construction of retrieval and processing buildings, excavation and retrieval of wastes from Pit 9, selective physical separation and chemical extraction, and stabilization of wastes either through thermal processing or by forming a stabilized concentrate. The proposed action would involve limited waste treatment process testing and full-scale waste treatment processing for cleaning up pre-1970 Transuranic (TRU) wastes in Pit 9. The purpose of this interim action is to expedite the overall cleanup at the RWMC and to reduce the risks associated with potential migration of Pit 9 wastes to the Snake River Plain Aquifer.

  9. Radioactive Waste Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1984-02-06

    To establish policies and guidelines by which the Department of Energy (DOE) manages tis radioactive waste, waste byproducts, and radioactively contaminated surplus facilities.

  10. Waste Treatment Plant Overview

    Office of Environmental Management (EM)

    contracted Bechtel National, Inc., to design and build the world's largest radioactive waste treatment plant. The Waste Treatment and Immobilization Plant (WTP), also known as the...

  11. Salt Waste Processing Initiatives

    Office of Environmental Management (EM)

    Patricia Suggs Salt Processing Team Lead Assistant Manager for Waste Disposition Project Office of Environmental Management Savannah River Site Salt Waste Processing Initiatives 2...

  12. Hanford Tank Waste Retrieval,

    Office of Environmental Management (EM)

    Tank Waste Retrieval, Treatment, and Disposition Framework September 24, 2013 U.S. Department of Energy Washington, D.C. 20585 Hanford Tank Waste Retrieval, Treatment, and...

  13. Transuranic Waste Requirements

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09

    The guide provides criteria for determining if a waste is to be managed in accordance with DOE M 435.1-1, Chapter III, Transuranic Waste Requirements.

  14. Waste-to-Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    into renewable energy, thereby enabling a national network of distributed power and biofuel production sites. Image courtesy of Iona Capital Waste-to-Energy Cycle Waste...

  15. The incorporation of P, S, Cr, F, Cl, I, Mn, Ti, U, and Bi into simulated nuclear waste glasses: Literature study

    SciTech Connect (OSTI)

    Langowski, M.H.

    1996-02-01

    Waste currently stored on the Hanford Reservation in underground tanks will be into High Level Waste (HLW) and Low Level Waste (LLW). The HLW melter will high-level and transuranic wastes to a vitrified form for disposal in a geological repository. The LLW melter will vitrify the low-level waste which is mainly a sodium solution. Characterization of the tank wastes is still in progress, and the pretreatment processes are still under development Apart from tank-to-tank variations, the feed delivered to the HLW melter will be subject to process control variability which consists of blending and pretreating the waste. The challenge is then to develop glass formulation models which can produce durable and processable glass compositions for all potential vitrification feed compositions and processing conditions. The work under HLW glass formulation is to study and model glass and melt pro functions of glass composition and temperature. The properties of interest include viscosity, electrical conductivity, liquidus temperature, crystallization, immiscibility durability. It is these properties that determine the glass processability and ac waste glass. Apart from composition, some properties, such as viscosity are affected by temperature. The processing temperature may vary from 1050{degrees}C to 1550{degrees}C dependent upon the melter type. The glass will also experience a temperature profile upon cooling. The purpose of this letter report is to assess the expected vitrification feed compositions for critical components with the greatest potential impact on waste loading for double shell tank (DST) and single shell tank (SST) wastes. The basis for critical component selection is identified along with the planned approach for evaluation. The proposed experimental work is a crucial part of model development and verification.

  16. DOE Chooses Idaho Treatment Group, LLC to Disposition Waste at the Advanced Mixed Waste Treatment Project: Contract will continue cleanup and waste operations at the Idaho Site

    Broader source: Energy.gov [DOE]

    Idaho Falls – In order to further meet the U.S. Department of Energy’s commitments to the citizens of the state of Idaho, the DOE today announced that it has selected Idaho Treatment Group, LLC (ITG) to perform waste processing at the Advanced Mixed Waste Treatment Project (AMWTP) at DOE’s Idaho Site near Idaho Falls.

  17. FLUIDIZED BED STEAM REFORMING ENABLING ORGANIC HIGH LEVEL WASTE DISPOSAL

    SciTech Connect (OSTI)

    Williams, M

    2008-05-09

    Waste streams planned for generation by the Global Nuclear Energy Partnership (GNEP) and existing radioactive High Level Waste (HLW) streams containing organic compounds such as the Tank 48H waste stream at Savannah River Site have completed simulant and radioactive testing, respectfully, by Savannah River National Laboratory (SRNL). GNEP waste streams will include up to 53 wt% organic compounds and nitrates up to 56 wt%. Decomposition of high nitrate streams requires reducing conditions, e.g. provided by organic additives such as sugar or coal, to reduce NOX in the off-gas to N2 to meet Clean Air Act (CAA) standards during processing. Thus, organics will be present during the waste form stabilization process regardless of the GNEP processes utilized and exists in some of the high level radioactive waste tanks at Savannah River Site and Hanford Tank Farms, e.g. organics in the feed or organics used for nitrate destruction. Waste streams containing high organic concentrations cannot be stabilized with the existing HLW Best Developed Available Technology (BDAT) which is HLW vitrification (HLVIT) unless the organics are removed by pretreatment. The alternative waste stabilization pretreatment process of Fluidized Bed Steam Reforming (FBSR) operates at moderate temperatures (650-750 C) compared to vitrification (1150-1300 C). The FBSR process has been demonstrated on GNEP simulated waste and radioactive waste containing high organics from Tank 48H to convert organics to CAA compliant gases, create no secondary liquid waste streams and create a stable mineral waste form.

  18. Method of forming a chemical composition

    DOE Patents [OSTI]

    Bingham, Dennis N. (Idaho Falls, ID); Wilding, Bruce M. (Idaho Falls, ID); Klingler, Kerry M. (Idaho Falls, ID); Zollinger, William T. (Idaho Falls, ID); Wendt, Kraig M. (Idaho Falls, ID)

    2007-10-09

    A method of forming a chemical composition such as a chemical hydride is described and which includes the steps of selecting a composition having chemical bonds and which is capable of forming a chemical hydride; providing a source of hydrogen; and exposing the selected composition to an amount of ionizing radiation to encourage the changing of the chemical bonds of the selected composition, and chemically reacting the selected composition with the source of hydrogen to facilitate the formation of a chemical hydride.

  19. Apparatus for incinerating hazardous waste

    DOE Patents [OSTI]

    Chang, Robert C. W. (Martinez, GA)

    1994-01-01

    An apparatus for incinerating wastes, including an incinerator having a combustion chamber, a fluidtight shell enclosing the combustion chamber, an afterburner, an off-gas particulate removal system and an emergency off-gas cooling system. The region between the inner surface of the shell and the outer surface of the combustion chamber forms a cavity. Air is supplied to the cavity and heated as it passes over the outer surface of the combustion chamber. Heated air is drawn from the cavity and mixed with fuel for input into the combustion chamber. The pressure in the cavity is maintained at least approximately 2.5 cm WC (about 1" WC) higher than the pressure in the combustion chamber. Gases cannot leak from the combustion chamber since the pressure outside the chamber (inside the cavity) is higher than the pressure inside the chamber. The apparatus can be used to treat any combustible wastes, including biological wastes, toxic materials, low level radioactive wastes, and mixed hazardous and low level transuranic wastes.

  20. Apparatus for incinerating hazardous waste

    DOE Patents [OSTI]

    Chang, R.C.W.

    1994-12-20

    An apparatus is described for incinerating wastes, including an incinerator having a combustion chamber, a fluid-tight shell enclosing the combustion chamber, an afterburner, an off-gas particulate removal system and an emergency off-gas cooling system. The region between the inner surface of the shell and the outer surface of the combustion chamber forms a cavity. Air is supplied to the cavity and heated as it passes over the outer surface of the combustion chamber. Heated air is drawn from the cavity and mixed with fuel for input into the combustion chamber. The pressure in the cavity is maintained at least approximately 2.5 cm WC higher than the pressure in the combustion chamber. Gases cannot leak from the combustion chamber since the pressure outside the chamber (inside the cavity) is higher than the pressure inside the chamber. The apparatus can be used to treat any combustible wastes, including biological wastes, toxic materials, low level radioactive wastes, and mixed hazardous and low level transuranic wastes. 1 figure.

  1. Hanford Site annual dangerous waste report: Volume 4, Waste Management Facility report, Radioactive mixed waste

    SciTech Connect (OSTI)

    1994-12-31

    This report contains information on radioactive mixed wastes at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, handling method and containment vessel, waste number, waste designation and amount of waste.

  2. Hanford Site annual dangerous waste report: Volume 2, Generator dangerous waste report, radioactive mixed waste

    SciTech Connect (OSTI)

    1994-12-31

    This report contains information on radioactive mixed wastes at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, waste number, waste designation, weight, and waste designation.

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

    SciTech Connect (OSTI)

    Amoroso, J.; Marra, J.

    2014-10-02

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

  4. Multi-step process for concentrating magnetic particles in waste sludges

    DOE Patents [OSTI]

    Watson, J.L.

    1990-07-10

    This invention involves a multi-step, multi-force process for dewatering sludges which have high concentrations of magnetic particles, such as waste sludges generated during steelmaking. This series of processing steps involves (1) mixing a chemical flocculating agent with the sludge; (2) allowing the particles to aggregate under non-turbulent conditions; (3) subjecting the mixture to a magnetic field which will pull the magnetic aggregates in a selected direction, causing them to form a compacted sludge; (4) preferably, decanting the clarified liquid from the compacted sludge; and (5) using filtration to convert the compacted sludge into a cake having a very high solids content. Steps 2 and 3 should be performed simultaneously. This reduces the treatment time and increases the extent of flocculation and the effectiveness of the process. As partially formed aggregates with active flocculating groups are pulled through the mixture by the magnetic field, they will contact other particles and form larger aggregates. This process can increase the solids concentration of steelmaking sludges in an efficient and economic manner, thereby accomplishing either of two goals: (a) it can convert hazardous wastes into economic resources for recycling as furnace feed material, or (b) it can dramatically reduce the volume of waste material which must be disposed. 7 figs.

  5. Multi-step process for concentrating magnetic particles in waste sludges

    DOE Patents [OSTI]

    Watson, John L. (Rolla, MO)

    1990-01-01

    This invention involves a multi-step, multi-force process for dewatering sludges which have high concentrations of magnetic particles, such as waste sludges generated during steelmaking. This series of processing steps involves (1) mixing a chemical flocculating agent with the sludge; (2) allowing the particles to aggregate under non-turbulent conditions; (3) subjecting the mixture to a magnetic field which will pull the magnetic aggregates in a selected direction, causing them to form a compacted sludge; (4) preferably, decanting the clarified liquid from the compacted sludge; and (5) using filtration to convert the compacted sludge into a cake having a very high solids content. Steps 2 and 3 should be performed simultaneously. This reduces the treatment time and increases the extent of flocculation and the effectiveness of the process. As partially formed aggregates with active flocculating groups are pulled through the mixture by the magnetic field, they will contact other particles and form larger aggregates. This process can increase the solids concentration of steelmaking sludges in an efficient and economic manner, thereby accomplishing either of two goals: (a) it can convert hazardous wastes into economic resources for recycling as furnace feed material, or (b) it can dramatically reduce the volume of waste material which must be disposed.

  6. CHEMICAL ANALYSIS OF SIMULATED HIGH LEVEL WASTE GLASSES TO SUPPORT SULFATE SOLUBILITY MODELING

    SciTech Connect (OSTI)

    Fox, K.; Marra, J.

    2014-08-14

    The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms both within the DOE complex and to some extent at U.K. sites. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerated cleanup missions. Much of the previous work on improving sulfate retention in waste glasses has been done on an empirical basis, making it difficult to apply the findings to future waste compositions despite the large number of glass systems studied. A more fundamental, rather than empirical, model of sulfate solubility in glass, under development at Sheffield Hallam University (SHU), could provide a solution to the issues of sulfate solubility. The model uses the normalized cation field strength index as a function of glass composition to predict sulfate capacity, and has shown early success for some glass systems. The objective of the current scope is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DOE waste vitrification efforts, allowing for enhanced waste loadings and waste throughput. A series of targeted glass compositions was selected to resolve data gaps in the current model. SHU fabricated these glasses and sent samples to the Savannah River National Laboratory (SRNL) for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for simulated waste glasses fabricated SHU in support of sulfate solubility model development. A review of the measured compositions revealed that there are issues with the B{sub 2}O{sub 3} and Fe{sub 2}O{sub 3} concentrations missing their targeted values by a significant amount for several of the study glasses. SHU is reviewing the fabrication of these glasses and the chemicals used in batching them to identify the source of these issues. The measured sulfate concentrations were all below their targeted values. This is expected, as the targeted concentrations likely exceeded the solubility limit for sulfate in these glass compositions. Some volatilization of sulfate may also have occurred during fabrication of the glasses. Measurements of the other oxides in the study glasses were reasonably close to their targeted values

  7. HAZARDOUS WASTE MANAGEMENT REFERENCE

    E-Print Network [OSTI]

    Winfree, Erik

    HAZARDOUS WASTE MANAGEMENT REFERENCE GUIDE Prepared by Environment, Health and Safety Office@caltech.edu http://safety.caltech.edu #12;Hazardous Waste Management Reference Guide Page 2 of 36 TABLE OF CONTENTS Satellite Accumulation Area 9 Waste Accumulation Facility 10 HAZARDOUS WASTE CONTAINER MANAGEMENT Labeling

  8. Robust Solution to Difficult Hydrogen Issues When Shipping Transuranic Waste to the Waste Isolation Pilot Plant

    SciTech Connect (OSTI)

    Countiss, S. S.; Basabilvazo, G. T.; Moody, D. C. III; Lott, S. A.; Pickerell, M.; Baca, T.; CH2M Hill; Tujague, S.; Svetlik, H.; Hannah, T.

    2003-02-27

    The Waste Isolation Pilot Plant (WIPP) has been open, receiving, and disposing of transuranic (TRU) waste since March 26, 1999. The majority of the waste has a path forward for shipment to and disposal at the WIPP, but there are about two percent (2%) or approximately 3,020 cubic meters (m{sup 3}) of the volume of TRU waste (high wattage TRU waste) that is not shippable because of gas generation limits set by the U.S. Nuclear Regulatory Commission (NRC). This waste includes plutonium-238 waste, solidified organic waste, and other high plutonium-239 wastes. Flammable gases are potentially generated during transport of TRU waste by the radiolysis of hydrogenous materials and therefore, the concentration at the end of the shipping period must be predicted. Two options are currently available to TRU waste sites for solving this problem: (1) gas generation testing on each drum, and (2) waste form modification by repackaging and/or treatment. Repackaging some of the high wattage waste may require up to 20:1 drum increase to meet the gas generation limits of less than five percent (5%) hydrogen in the inner most layer of confinement (the layer closest to the waste). (This is the limit set by the NRC.) These options increase waste handling and transportation risks and there are high costs and potential worker exposure associated with repackaging this high-wattage TRU waste. The U.S. Department of Energy (DOE)'s Carlsbad Field Office (CBFO) is pursuing a twofold approach to develop a shipping path for these wastes. They are: regulatory change and technology development. For the regulatory change, a more detailed knowledge of the high wattage waste (e.g., void volumes, gas generation potential of specific chemical constituents) may allow refinement of the current assumptions in the gas generation model for Safety Analysis Reports for Packaging for Contact-Handled (CH) TRU waste. For technology development, one of the options being pursued is the use of a robust container, the ARROW-PAK{trademark} System. (1) The ARROW-PAK{trademark} is a macroencapsulation treatment technology, developed by Boh Environmental, LLC, New Orleans, Louisiana. This technology has been designed to withstand any unexpected hydrogen deflagration (i.e. no consequence) and other benefits such as criticality control.

  9. The effects of ocean-dumped wastes on marine phytoplankto 

    E-Print Network [OSTI]

    Schwab, Claude Raymond

    1980-01-01

    even such hazardous and persistent wastes as chemical warfare agents and radioactive wastes. The status of ocean dumping at this time is summarized in an excellent report by Smith and Brown (1971). Much of this same information, with some updates... phytoplankton so that this information can be used to evaluate the practice of ocean dumping in general and, more specifically, for the selected wastes studied, These studies will use radioactive carbon assimilation and adenosine triphosphate measurements so...

  10. Process to separate transuranic elements from nuclear waste

    DOE Patents [OSTI]

    Johnson, Terry R. (Wheaton, IL); Ackerman, John P. (Downers Grove, IL); Tomczuk, Zygmunt (Orland Park, IL); Fischer, Donald F. (Glen Ellyn, IL)

    1989-01-01

    A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

  11. Thermal and chemical remediation of mixed waste

    DOE Patents [OSTI]

    Nelson, Paul A. (Wheaton, IL); Swift, William M. (Downers Grove, IL)

    1994-01-01

    A process and system for treating organic waste materials without venting gaseous emissions to the atmosphere. A fluidized bed including lime particles is operated at a temperature of at least 500.degree. C. by blowing gas having 20%/70% oxygen upwardly through the bed particles at a rate sufficient to fluidize same. A toxic organic waste material is fed into the fluidized bed where the organic waste material reacts with the lime forming CaCO.sub.3. The off gases are filtered and cooled to condense water which is separated. A portion of the calcium carbonate formed during operation of the fluidized bed is replaced with lime particles. The off gases from the fluidized bed after drying are recirculated until the toxic organic waste material in the bed is destroyed.

  12. Thermal and chemical remediation of mixed waste

    DOE Patents [OSTI]

    Nelson, P.A.; Swift, W.M.

    1994-08-09

    A process and system for treating organic waste materials without venting gaseous emissions to the atmosphere. A fluidized bed including lime particles is operated at a temperature of at least 500 C by blowing gas having 20%/70% oxygen upwardly through the bed particles at a rate sufficient to fluidize same. A toxic organic waste material is fed into the fluidized bed where the organic waste material reacts with the lime forming CaCO[sub 3]. The off gases are filtered and cooled to condense water which is separated. A portion of the calcium carbonate formed during operation of the fluidized bed is replaced with lime particles. The off gases from the fluidized bed after drying are recirculated until the toxic organic waste material in the bed is destroyed. 3 figs.

  13. Thermal denitration and mineralization of waste constituents

    SciTech Connect (OSTI)

    Nenni, J.A.; Boardman, R.D.

    1997-08-01

    In order to produce a quality grout from LLW using hydraulic cements, proper conditioning of the waste is essential for complete cement curing. Several technologies were investigated as options for conditions. Since the LLW is dilute, removal of all, or most, of the water will significantly reduce the final waste volume. Neutralization of the LLW is also desirable since acidic liquids to not allow cement to cure properly. The nitrate compounds are very soluble and easily leached from solid waste forms; therefore, denitration is desirable. Thermal and chemical denitration technologies have the advantages of water removal, neutralization, and denitration. The inclusion of additives during thermal treatment were investigated as a method of forming insoluable waste conditions.

  14. Cold bond agglomeration of waste oxides for recycling

    SciTech Connect (OSTI)

    D`Alessio, G.; Lu, W.K. [McMaster Univ., Hamilton, Ontario (Canada). Dept. of Materials Science and Engineering

    1996-12-31

    Recycling of waste oxides has been an on-going challenge for integrated steel plants. The majority of these waste oxides are collected from the cleaning systems of ironmaking and steelmaking processes, and are usually in the form of fine particulates and slurries. In most cases, these waste materials are contaminated by oils and heavy metals and often require treatment at a considerable expense prior to landfill disposal. This contamination also limits the re-use or recycling potential of these oxides as secondary resources of reliable quality. However, recycling of some selected wastes in blast furnaces or steelmaking vessels is possible, but first requires agglomeration of the fine particulate by such methods as cold bond briquetting. Cold bond briquetting technology provides both mechanical compacting and bonding (with appropriate binders) of the particulates. This method of recycling has the potential to be economically viable and environmentally sustainable. The nature of the present study is cold bond briquetting of iron ore pellet fines with a molasses-cement-H{sub 2}O binder for recycling in a blast furnace. The inclusion of molasses is for its contribution to the green strength of briquettes. During the curing stage, significant gains in strength may be credited to molasses in the presence of cement. The interactions of cement (and its substitutes), water and molasses and their effects on the properties of the agglomerates during and after various curing conditions were investigated. Tensile strengths of briquettes made in the laboratory and subjected to experimental conditions which simulated the top part of a blast furnace shaft were also examined.

  15. Waste Package Lifting Calculation

    SciTech Connect (OSTI)

    H. Marr

    2000-05-11

    The objective of this calculation is to evaluate the structural response of the waste package during the horizontal and vertical lifting operations in order to support the waste package lifting feature design. The scope of this calculation includes the evaluation of the 21 PWR UCF (pressurized water reactor uncanistered fuel) waste package, naval waste package, 5 DHLW/DOE SNF (defense high-level waste/Department of Energy spent nuclear fuel)--short waste package, and 44 BWR (boiling water reactor) UCF waste package. Procedure AP-3.12Q, Revision 0, ICN 0, calculations, is used to develop and document this calculation.

  16. The siting record: An account of the programs of federal agencies and events that have led to the selection of a potential site for a geologic respository for high-level radioactive waste

    SciTech Connect (OSTI)

    Lomenick, T.F.

    1996-03-01

    This record of siting a geologic repository for high-level radioactive wastes (HLW) and spent fuel describes the many investigations that culminated on December 22, 1987 in the designation of Yucca Mountain (YM), as the site to undergo detailed geologic characterization. It recounts the important issues and events that have been instrumental in shaping the course of siting over the last three and one half decades. In this long task, which was initiated in 1954, more than 60 regions, areas, or sites involving nine different rock types have been investigated. This effort became sharply focused in 1983 with the identification of nine potentially suitable sites for the first repository. From these nine sites, five were subsequently nominated by the U.S. Department of Energy (DOE) as suitable for characterization and then, in 1986, as required by the Nuclear Waste Policy Act of 1982 (NWPA), three of these five were recommended to the President as candidates for site characterization. President Reagan approved the recommendation on May 28, 1986. DOE was preparing site characterization plans for the three candidate sites, namely Deaf Smith County, Texas; Hanford Site, Washington; and YM. As a consequence of the 1987 Amendment to the NWPA, only the latter was authorized to undergo detailed characterization. A final Site Characterization Plan for Yucca Mountain was published in 1988. Prior to 1954, there was no program for the siting of disposal facilities for high-level waste (HLW). In the 1940s and 1950s, the volume of waste, which was small and which resulted entirely from military weapons and research programs, was stored as a liquid in large steel tanks buried at geographically remote government installations principally in Washington and Tennessee.

  17. HANFORD MEDIUM & LOW CURIE WASTE PRETREATMENT PROJECT PHASE 1 LAB REPORT

    SciTech Connect (OSTI)

    HAMILTON, D.W.

    2006-01-30

    A fractional crystallization (FC) process is being developed to supplement tank waste pretreatment capabilities provided by the Waste Treatment and Immobilization Plant (WTP). FC can process many tank wastes, separating wastes into a low-activity fraction (LAW) and high-activity fraction (HLW). The low-activity fraction can be immobilized in a glass waste form by processing in the bulk vitrification (BV) system.

  18. Understanding radioactive waste

    SciTech Connect (OSTI)

    Murray, R.L.

    1981-12-01

    This document contains information on all aspects of radioactive wastes. Facts are presented about radioactive wastes simply, clearly and in an unbiased manner which makes the information readily accessible to the interested public. The contents are as follows: questions and concerns about wastes; atoms and chemistry; radioactivity; kinds of radiation; biological effects of radiation; radiation standards and protection; fission and fission products; the Manhattan Project; defense and development; uses of isotopes and radiation; classification of wastes; spent fuels from nuclear reactors; storage of spent fuel; reprocessing, recycling, and resources; uranium mill tailings; low-level wastes; transportation; methods of handling high-level nuclear wastes; project salt vault; multiple barrier approach; research on waste isolation; legal requiremnts; the national waste management program; societal aspects of radioactive wastes; perspectives; glossary; appendix A (scientific American articles); appendix B (reference material on wastes). (ATT)

  19. New Innovations in Highly Ion Specific Media for Recalcitrant Waste stream Radioisotopes

    SciTech Connect (OSTI)

    Denton, M. S.; Wilson, J.; Ahrendt, M. [RWE NUKEM Corporation (RNC), 800 Oak Ridge Tnpk., Suite A701, Oak Ridge, TN 37830 (United States); Bostick, W. D. [Materials and Chemistry Laboratory (MCL), Inc., East Tennessee Technology Park, Building K-1006, 2010 Highway 58, Suite 1000, Oak Ridge, TN 37830 (United States); DeSilva, F.; Meyers, P. [ResinTech, Inc., 1 ResinTech Plaza, 160 Cooper Road, West Berlin, NJ 08091 (United States)

    2006-07-01

    Specialty ion specific media were examined and developed for, not only pre- and post-outage waste streams, but also for very difficult outage waste streams. This work was carried out on first surrogate waste streams, then laboratory samples of actual waste streams, and, finally, actual on-site waste streams. This study was particularly focused on PWR wastewaters such as Floor Drain Tank (FDT), Boron Waste Storage Tank (BWST), and Waste Treatment Tank (WTT, or discharge tank). Over the last half decade, or so, treatment technologies have so greatly improved and discharge levels have become so low, that certain particularly problematic isotopes, recalcitrant to current treatment skids, are all that remain prior to discharge. In reality, they have always been present, but overshadowed by the more prevalent and higher activity isotopes. Such recalcitrants include cobalt, especially Co 58 [both ionic/soluble (total dissolved solids, TDS) and colloidal (total suspended solids, TSS)] and antimony (Sb). The former is present in most FDT and BWST wastewaters, while the Sb is primarily present in BWST waste streams. The reasons Co 58 can be elusive to granulated activated carbon (GAC), ultrafiltration (UF) and ion exchange (IX) demineralizers is that it forms submicron colloids as well as has a tendency to form metal complexes with chelating agents (e.g., ethylene diamine tetraacetic acid, or EDTA). Such colloids and non-charged complexes will pass through the entire treatment skid. Antimony (Sb) on the other hand, has little or no ionic charge, and will, likewise, pass through both the filtration and de-min skids into the discharge tanks. While the latter will sometimes (the anionic vs. the cationic or neutral species) be removed on the anion bed(s), it will slough off (snow-plow effect) when a higher affinity anion (iodine slugs, etc.) comes along; thus causing effluents not meeting discharge criteria. The answer to these problems found in this study, during an actual Nuclear Power Plant (NPP) outage cycle and recovery (four months), was the down-select and development of a number of highly ion specific media for the specific removal of such elusive isotopes. Over three dozen media including standard cation and anion ion exchangers, specialty IX, standard carbons, and, finally, chemically doped media (e.g., carbon and alumina substrates). The latter involved doping with iron, manganese, and even metals. The media down-select was carried out on actual plant waste streams so that all possible outage affects were accounted for, and distribution coefficients (Kd's) were determined (vs. decontamination factors, DF's, or percent removals). Such Kd's, in milliliters of solution per gram of media (mug), produce data indicative of the longevity of the media in that particular waste stream. Herein, the down-select is reported in Pareto (decreasing order) tables. Further affects such as the presence of high cobalt concentrations, high boron concentrations, the presence of hydrazine and chelating agents, and extreme pH conditions. Of particular importance here is to avoid the affinity of competing ions (e.g., a Sb specific media having more than a slight affinity for Co). The latter results in the snow-plow effect of sloughing off 3 to 4 times the cobalt into the effluent as was in the feed upon picking up the Sb. The study was quite successful and resulted in the development of and selection of a resin-type and two granular media for antimony removal, and two resin-types and a granular media for cobalt removal. The decontamination factors for both media were hundreds to thousands of times that of the full filtration and de-min. (authors)

  20. Spectroelectrochemical Sensing Based on Multimode Selectivity Simultaneously Achievable in a Single Device. 11. Design and Evaluation of a Small Portable Sensor for the Determination of Ferrocyanide in Hanford Waste Samples.

    SciTech Connect (OSTI)

    Stegemiller, Michael L.; Heineman, William R.; Seliskar, Carl J.; Ridgeway, Thomas H.; Bryan, Samuel A.; Hubler, Timothy L.; Sell, Rachel L.

    2003-01-01

    A portable spectroelectrochemical sensor has been designed, evaluated, and demonstrated on a complex sample of radioactive waste. The sensor consisted of a black delrin sample compartment with a total internal sample volume of 800 ?L, attached to an indium tin oxide coated glass multiple internal reflection optical element. Detection was by total internal reflection of light from a blue light emitting diode source. After a 10 min uptake for each standard, the sensor showed a linear response in absorbance change for 5 x 10-5 to 5 x 10-3 M ferrocyanide with electrochemical modulation by scanning at 25 mV/s from ? 0.30 V to + 0.55 V vs. a Ag/AgCl reference electrode. Due to the complex nature of Hanford radioactive tank waste samples containing ferrocyanide, a standard addition method was developed for analysis. The spectroelectrochemical sensor determined a concentration of 9.2 mM ferrocyanide for U-Plant-2 simulant solution containing 9.38 mM ferrocyanide that was prepared according to Hanford process flowsheets. A radioactive tank waste sample from Hanford Tank 241-C-112 was determined to be 1.0 mM in ferrocyanide using the spectroelectrochemical sensor. The accepted value for ferrocyanide concentration in the sample is 0.61 mM as determined by IR spectroscopy.

  1. Guidance manual for the identification of hazardous wastes delivered to publicly owned treatment works by truck, rail, or dedicated pipe

    SciTech Connect (OSTI)

    Not Available

    1987-06-01

    The manual is directed towards two types of facilities: First, guidance is to POTWs that wish to preclude the entry of hazardous wastes into their facilities and avoid regulation and liability under RCRA. Administrative/technical recommendations for control of such wastes is provided, many of which are already in use by POTWs. Second, the responsibilities of POTWs that choose to accept hazardous wastes from truck, rail, or dedicated pipeline are discussed, including relevant regulatory provisions, strict liability and corrective action requirements for releases, and recommended procedures for waste acceptance/management. The manual describes the RCRA regulatory status of wastes that POTW operators typically may encounter. The manual includes a Waste Monitoring Plan. Appendices give the following: RCRA lists; RCRA listed hazardous wastes; examples of POTW sewer use ordinance language, waste hauler permit; waste tracking form, notification of hazardous waste activity; uniform hazardous waste manifest; biennial hazardous waste report; and state hazardous waste contacts.

  2. MANAGING HANFORD'S LEGACY NO-PATH-FORWARD WASTES TO DISPOSITION

    SciTech Connect (OSTI)

    WEST LD

    2011-01-13

    The U.S. Department of Energy (DOE) Richland Operations Office (RL) has adopted the 2015 Vision for Cleanup of the Hanford Site. This vision will protect the Columbia River, reduce the Site footprint, and reduce Site mortgage costs. The CH2M HILL Plateau Remediation Company's (CHPRC) Waste and Fuels Management Project (W&FMP) and their partners support this mission by providing centralized waste management services for the Hanford Site waste generating organizations. At the time of the CHPRC contract award (August 2008) slightly more than 9,000 m{sup 3} of waste was defined as 'no-path-forward waste.' The majority of these wastes are suspect transuranic mixed (TRUM) wastes which are currently stored in the low-level Burial Grounds (LLBG), or stored above ground in the Central Waste Complex (CWC). A portion of the waste will be generated during ongoing and future site cleanup activities. The DOE-RL and CHPRC have collaborated to identify and deliver safe, cost-effective disposition paths for 90% ({approx}8,000 m{sup 3}) of these problematic wastes. These paths include accelerated disposition through expanded use of offsite treatment capabilities. Disposal paths were selected that minimize the need to develop new technologies, minimize the need for new, on-site capabilities, and accelerate shipments of transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico.

  3. Hanford's Simulated Low Activity Waste Cast Stone Processing

    SciTech Connect (OSTI)

    Kim, Young

    2013-08-20

    Cast Stone is undergoing evaluation as the supplemental treatment technology for Hanford’s (Washington) high activity waste (HAW) and low activity waste (LAW). This report will only cover the LAW Cast Stone. The programs used for this simulated Cast Stone were gradient density change, compressive strength, and salt waste form phase identification. Gradient density changes show a favorable outcome by showing uniformity even though it was hypothesized differently. Compressive strength exceeded the minimum strength required by Hanford and greater compressive strength increase seen between the uses of different salt solution The salt waste form phase is still an ongoing process as this time and could not be concluded.

  4. Thermal Predictions of the Cooling of Waste Glass Canisters

    SciTech Connect (OSTI)

    Donna Post Guillen

    2014-11-01

    Radioactive liquid waste from five decades of weapons production is slated for vitrification at the Hanford site. The waste will be mixed with glass forming additives and heated to a high temperature, then poured into canisters within a pour cave where the glass will cool and solidify into a stable waste form for disposal. Computer simulations were performed to predict the heat rejected from the canisters and the temperatures within the glass during cooling. Four different waste glass compositions with different thermophysical properties were evaluated. Canister centerline temperatures and the total amount of heat transfer from the canisters to the surrounding air are reported.

  5. Sodium-Bearing Waste Treatment Alternatives Implementation Study

    SciTech Connect (OSTI)

    Charles M. Barnes; James B. Bosley; Clifford W. Olsen

    2004-07-01

    The purpose of this document is to discuss issues related to the implementation of each of the five down-selected INEEL/INTEC radioactive liquid waste (sodium-bearing waste - SBW) treatment alternatives and summarize information in three main areas of concern: process/technical, environmental permitting, and schedule. Major implementation options for each treatment alternative are also identified and briefly discussed. This report may touch upon, but purposely does not address in detail, issues that are programmatic in nature. Examples of these include how the SBW will be classified with respect to the Nuclear Waste Policy Act (NWPA), status of Waste Isolation Pilot Plant (WIPP) permits and waste storage availability, available funding for implementation, stakeholder issues, and State of Idaho Settlement Agreement milestones. It is assumed in this report that the SBW would be classified as a transuranic (TRU) waste suitable for disposal at WIPP, located in New Mexico, after appropriate treatment to meet transportation requirements and waste acceptance criteria (WAC).

  6. Radioactive Waste Management Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09

    This Manual further describes the requirements and establishes specific responsibilities for implementing DOE O 435.1, Radioactive Waste Management, for the management of DOE high-level waste, transuranic waste, low-level waste, and the radioactive component of mixed waste. Change 1 dated 6/19/01 removes the requirement that Headquarters is to be notified and the Office of Environment, Safety and Health consulted for exemptions for use of non-DOE treatment facilities. Certified 1-9-07.

  7. Radiation Effects in Nuclear Waste Materials

    SciTech Connect (OSTI)

    Weber, William J.

    2005-09-30

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

  8. Radiation Effects in Nuclear Waste Materials

    SciTech Connect (OSTI)

    Weber, William J.

    2005-06-01

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

  9. Microwave solidification development for Rocky Flats waste

    SciTech Connect (OSTI)

    Dixon, D.; Erle, R.; Eschen, V.

    1994-04-01

    The Microwave Engineering Team at the Rocky Flats Plant has developed a production-scale system for the treatment of hazardous, radioactive, and mixed wastes using microwave energy. The system produces a vitreous final form which meets the acceptance criteria for shipment and disposal. The technology also has potential for application on various other waste streams from the public and private sectors. Technology transfer opportunities are being identified and pursued for commercialization of the microwave solidification technology.

  10. Vitrification of organics-containing wastes

    DOE Patents [OSTI]

    Bickford, D.F.

    1995-01-01

    A process for stabilizing organics-containing waste materials and recovery metals therefrom, and a waste glass product made according to the process are described. Vitrification of wastes such as organic ion exchange resins, electronic components and the like can be accomplished by mixing at least one transition metal oxide with the wastes, and, if needed, glass formers to compensate for a shortage of silicates or other glass formers in the wastes. The transition metal oxide increases the rate of oxidation of organic materials in the wastes to improve the composition of the glass-forming mixture: at low temperatures, the oxide catalyzes oxidation of a portion of the organics in the waste; at higher temperatures, the oxide dissolves and the resulting oxygen ions oxidize more of the organics; and at vitrification temperatures, the metal ions conduct oxygen into the melt to oxidize the remaining organics. In addition, the transition metal oxide buffers the redox potential of the glass melt so that metals such as Au, Pt, Ag, and Cu separate form the melt in the metallic state and can be recovered. After the metals are recovered, the remainder of the melt is allowed to cool and may subsequently be disposed of. The product has good leaching resistance and can be disposed of in an ordinary landfill, or, alternatively, used as a filler in materials such as concrete, asphalt, brick and tile.

  11. Chemical Stabilization of Hanford Tank Residual Waste

    SciTech Connect (OSTI)

    Cantrell, Kirk J.; Um, Wooyong; Williams, Benjamin D.; Bowden, Mark E.; Gartman, Brandy N.; Lukens, Wayne W.; Buck, Edgar C.; Mausolf, Edward J.

    2014-03-01

    Three different chemical treatment methods were tested for their ability to stabilize residual waste from Hanford tank C-202 for reducing contaminant release (Tc, Cr, and U in particular). The three treatment methods tested were lime addition [Ca(OH)2], an in-situ Ceramicrete waste form based on chemically bonded phosphate ceramics, and a ferrous iron/goethite treatment. These approaches rely on formation of insoluble forms of the contaminants of concern (lime addition and ceramicrete) and chemical reduction followed by co-precipitation (ferrous iron/goethite incorporation treatment). The results have demonstrated that release of the three most significant mobile contaminants of concern from tank residual wastes can be dramatically reduced after treatment compared to contact with simulated grout porewater without treatment. For uranium, all three treatments methods reduced the leachable uranium concentrations by well over three orders of magnitude. In the case of uranium and technetium, released concentrations were well below their respective MCLs for the wastes tested. For tank C-202 residual waste, chromium release concentrations were above the MCL but were considerably reduced relative to untreated tank waste. This innovative approach has the potential to revolutionize Hanford’s tank retrieval process, by allowing larger volumes of residual waste to be left in tanks while providing an acceptably low level of risk with respect to contaminant release that is protective of the environment and human health. Such an approach could enable DOE to realize significant cost savings through streamlined retrieval and closure operations.

  12. Characterization of the Defense Waste Processing Facility (DWPF) Environmental Assessment (EA) glass Standard Reference Material. Revision 1

    SciTech Connect (OSTI)

    Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Crawford, C.L.; Pickett, M.A.

    1993-06-01

    Liquid high-level nuclear waste at the Savannah River Site (SRS) will be immobilized by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Other waste form producers, such as West Valley Nuclear Services (WVNS) and the Hanford Waste Vitrification Project (HWVP), will also immobilize high-level radioactive waste in borosilicate glass. The canistered waste will be stored temporarily at each facility for eventual permanent disposal in a geologic repository. The Department of Energy has defined a set of requirements for the canistered waste forms, the Waste Acceptance Product Specifications (WAPS). The current Waste Acceptance Primary Specification (WAPS) 1.3, the product consistency specification, requires the waste form producers to demonstrate control of the consistency of the final waste form using a crushed glass durability test, the Product Consistency Test (PCI). In order to be acceptable, a waste glass must be more durable during PCT analysis than the waste glass identified in the DWPF Environmental Assessment (EA). In order to supply all the waste form producers with the same standard benchmark glass, 1000 pounds of the EA glass was fabricated. The chemical analyses and characterization of the benchmark EA glass are reported. This material is now available to act as a durability and/or redox Standard Reference Material (SRM) for all waste form producers.

  13. Waste reduction through consumer education. Final report

    SciTech Connect (OSTI)

    Harrison, E.Z.

    1996-05-01

    The Waste Reduction through Consumer Education research project was conducted to determine how environmental educational strategies influence purchasing behavior in the supermarket. The objectives were to develop, demonstrate, and evaluate consumer education strategies for waste reduction. The amount of waste generated by packaging size and form, with an adjustment for local recyclability of waste, was determined for 14 product categories identified as having more waste generating and less waste generating product choices (a total of 484 products). Using supermarket scan data and shopper identification numbers, the research tracked the purchases of shoppers in groups receiving different education treatments for 9 months. Statistical tests applied to the purchase data assessed patterns of change between the groups by treatment period. Analysis of the data revealed few meaningful statistical differences between study groups or changes in behavior over time. Findings suggest that broad brush consumer education about waste reduction is not effective in changing purchasing behaviors in the short term. However, it may help create a general awareness of the issues surrounding excess packaging and consumer responsibility. The study concludes that the answer to waste reduction in the future may be a combination of voluntary initiatives by manufacturers and retailers, governmental intervention, and better-informed consumers.

  14. Recovery of fissile materials from nuclear wastes

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1999-01-01

    A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, organic material and other carbon-containing material.

  15. Nevada Test Site Waste Acceptance Criteria (NTSWAC), Rev. 7-01

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2009-05-01

    This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, Nevada Test Site Waste Acceptance Criteria (NTSWAC). The NTSWAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NTSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex for disposal.

  16. Technical area status report for waste destruction and stabilization

    SciTech Connect (OSTI)

    Dalton, J.D.; Harris, T.L.; DeWitt, L.M.

    1993-08-01

    The Office of Environmental Restoration and Waste Management (EM) was established by the Department of Energy (DOE) to direct and coordinate waste management and site remediation programs/activities throughout the DOE complex. In order to successfully achieve the goal of properly managing waste and the cleanup of the DOE sites, the EM was divided into five organizations: the Office of Planning and Resource Management (EM-10); the Office of Environmental Quality Assurance and Resource Management (EM-20); the Office of Waste Operations (EM-30); the Office of Environmental Restoration (EM-40); and the Office of Technology and Development (EM-50). The mission of the Office of Technology Development (OTD) is to develop treatment technologies for DOE`s operational and environmental restoration wastes where current treatment technologies are inadequate or not available. The Mixed Waste Integrated Program (MWIP) was created by OTD to assist in the development of treatment technologies for the DOE mixed low-level wastes (MLLW). The MWIP has established five Technical Support Groups (TSGs) whose purpose is to identify, evaluate, and develop treatment technologies within five general technical areas representing waste treatment functions from initial waste handling through generation of final waste forms. These TSGs are: (1) Front-End Waste Handling, (2) Physical/Chemical Treatment, (3) Waste Destruction and Stabilization, (4) Second-Stage Destruction and Offgas Treatment, and (5) Final Waste Forms. This report describes the functions of the Waste Destruction and Stabilization (WDS) group. Specifically, the following items are discussed: DOE waste stream identification; summary of previous efforts; summary of WDS treatment technologies; currently funded WDS activities; and recommendations for future activities.

  17. Waste Management Quality Assurance Plan

    E-Print Network [OSTI]

    Waste Management Group

    2006-01-01

    LBNL/PUB-5352, Revision 6 Waste Management QualityAssurance Plan Waste Management Group Environment, HealthRev. 6 WM QA Plan Waste Management Quality Assurance Plan

  18. Response to West Cumbria MRWS consultation: Why a deep nuclear waste repository should not be

    E-Print Network [OSTI]

    Response to West Cumbria MRWS consultation: Why a deep nuclear waste repository should not be sited geological nuclear waste repository. There a suspicion of predetermination because the only district that has. National and international guidance on how best to select potential sites for deep geological nuclear waste

  19. Waste management project's alternatives: A risk-based multi-criteria assessment (RBMCA) approach

    SciTech Connect (OSTI)

    Karmperis, Athanasios C.; Sotirchos, Anastasios; Aravossis, Konstantinos; Tatsiopoulos, Ilias P.

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer We examine the evaluation of a waste management project's alternatives. Black-Right-Pointing-Pointer We present a novel risk-based multi-criteria assessment (RBMCA) approach. Black-Right-Pointing-Pointer In the RBMCA the evaluation criteria are based on the quantitative risk analysis of the project's alternatives. Black-Right-Pointing-Pointer Correlation between the criteria weight values and the decision makers' risk preferences is examined. Black-Right-Pointing-Pointer Preference to the multi-criteria against the one-criterion evaluation process is discussed. - Abstract: This paper examines the evaluation of a waste management project's alternatives through a quantitative risk analysis. Cost benefit analysis is a widely used method, in which the investments are mainly assessed through the calculation of their evaluation indicators, namely benefit/cost (B/C) ratios, as well as the quantification of their financial, technical, environmental and social risks. Herein, a novel approach in the form of risk-based multi-criteria assessment (RBMCA) is introduced, which can be used by decision makers, in order to select the optimum alternative of a waste management project. Specifically, decision makers use multiple criteria, which are based on the cumulative probability distribution functions of the alternatives' B/C ratios. The RBMCA system is used for the evaluation of a waste incineration project's alternatives, where the correlation between the criteria weight values and the decision makers' risk preferences is analyzed and useful conclusions are discussed.

  20. JLF Forms

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTechtail.Theory ofDid you notHeat Pumps Heat PumpsfacilityviaGasforVendorsThu, 07JLF Forms

  1. Request Form

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home RoomPreservation of Fe(II) by Carbon-RichProtonAboutNuclear NonproliferationRequestForm Sign In

  2. Request Form

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home RoomPreservation of Fe(II) by Carbon-RichProtonAboutNuclear NonproliferationRequestForm Sign

  3. High Level Waste System Impacts from Small Column Ion Exchange Implementation

    SciTech Connect (OSTI)

    McCabe, D. J.; Hamm, L. L.; Aleman, S. E.; Peeler, D. K.; Herman, C. C.; Edwards, T. B.

    2005-08-18

    The objective of this task is to identify potential waste streams that could be treated with the Small Column Ion Exchange (SCIX) and perform an initial assessment of the impact of doing so on the High-Level Waste (HLW) system. Design of the SCIX system has been performed as a backup technology for decontamination of High-Level Waste (HLW) at the Savannah River Site (SRS). The SCIX consists of three modules which can be placed in risers inside underground HLW storage tanks. The pump and filter module and the ion exchange module are used to filter and decontaminate the aqueous tank wastes for disposition in Saltstone. The ion exchange module contains Crystalline Silicotitanate (CST in its engineered granular form is referred to as IONSIV{reg_sign} IE-911), and is selective for removal of cesium ions. After the IE-911 is loaded with Cs-137, it is removed and the column is refilled with a fresh batch. The grinder module is used to size-reduce the cesium-loaded IE-911 to make it compatible with the sludge vitrification system in the Defense Waste Processing Facility (DWPF). If installed at the SRS, this SCIX would need to operate within the current constraints of the larger HLW storage, retrieval, treatment, and disposal system. Although the equipment has been physically designed to comply with system requirements, there is also a need to identify which waste streams could be treated, how it could be implemented in the tank farms, and when this system could be incorporated into the HLW flowsheet and planning. This document summarizes a preliminary examination of the tentative HLW retrieval plans, facility schedules, decontamination factor targets, and vitrified waste form compatibility, with recommendations for a more detailed study later. The examination was based upon four batches of salt solution from the currently planned disposition pathway to treatment in the SCIX. Because of differences in capabilities between the SRS baseline and SCIX, these four batches were combined into three batches for a total of about 3.2 million gallons of liquid waste. The chemical and radiological composition of these batches was estimated from the SpaceMan Plus{trademark} model using the same data set and assumptions as the baseline plans.

  4. Cast Stone Formulation for Nuclear Waste Immobilization at Higher Sodium Concentrations

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Fox, Kevin; Cozzi, Alex; Roberts, Kimberly; Edwards, Thomas

    2014-11-01

    Low activity radioactive waste at U.S. Department of Energy sites can be immobilized for permanent disposal using cementitious waste forms. This study evaluated waste forms produced with simulated wastes at concentrations up to twice that of currently operating processes. The simulated materials were evaluated for their fresh properties, which determine processability, and cured properties, which determine waste form performance. The results show potential for greatly reducing the volume of material. Fresh properties were sufficient to allow for processing via current practices. Cured properties such as compressive strength meet disposal requirements. Leachability indices provide an indication of expected long-term performance.

  5. Composition and process for the encapsulation and stabilization of radioactive, hazardous and mixed wastes

    DOE Patents [OSTI]

    Kalb, P.D.; Colombo, P.

    1998-03-24

    The present invention provides a composition and process for disposal of radioactive, hazardous and mixed wastes. The present invention preferably includes a process for multibarrier encapsulation of radioactive, hazardous and mixed wastes by combining substantially simultaneously dry waste powder, a non-biodegradable thermoplastic polymer and an anhydrous additive in an extruder to form a homogeneous molten matrix. The molten matrix may be directed in a ``clean`` polyethylene liner, allowed to cool, thus forming a monolithic waste form which provides a multibarrier to the dispersion of wastes into the environment. 2 figs.

  6. Composition and process for the encapsulation and stabilization of radioactive, hazardous and mixed wastes

    DOE Patents [OSTI]

    Kalb, Paul D. (Wading River, NY); Colombo, Peter (Patchogue, NY)

    1999-07-20

    The present invention provides a composition and process for disposal of radioactive, hazardous and mixed wastes. The present invention preferably includes a process for multibarrier encapsulation of radioactive, hazardous and mixed wastes by combining substantially simultaneously dry waste powder, a non-biodegradable thermoplastic polymer and an anhydrous additive in an extruder to form a homogenous molten matrix. The molten matrix may be directed in a "clean" polyethylene liner, allowed to cool, thus forming a monolithic waste form which provides a multibarrier to the dispersion of wastes into the environment.

  7. Composition and process for the encapsulation and stabilization of radioactive, hazardous and mixed wastes

    DOE Patents [OSTI]

    Kalb, Paul D. (Wading River, NY); Colombo, Peter (Patchogue, NY)

    1998-03-24

    The present invention provides a composition and process for disposal of radioactive, hazardous and mixed wastes. The present invention preferably includes a process for multibarrier encapsulation of radioactive, hazardous and mixed wastes by combining substantially simultaneously dry waste powder, a non-biodegradable thermoplastic polymer and an anhydrous additive in an extruder to form a homogenous molten matrix. The molten matrix may be directed in a "clean" polyethylene liner, allowed to cool, thus forming a monolithic waste form which provides a multibarrier to the dispersion of wastes into the environment.

  8. Composition and process for the encapsulation and stabilization of radioactive hazardous and mixed wastes

    DOE Patents [OSTI]

    Kalb, P.D.; Colombo, P.

    1997-07-15

    The present invention provides a composition and process for disposal of radioactive, hazardous and mixed wastes. The present invention preferably includes a process for multibarrier encapsulation of radioactive, hazardous and mixed wastes by combining substantially simultaneously dry waste powder, a non-biodegradable thermoplastic polymer and an anhydrous additive in an extruder to form a homogeneous molten matrix. The molten matrix may be directed in a ``clean`` polyethylene liner, allowed to cool, thus forming a monolithic waste form which provides a multibarrier to the dispersion of wastes into the environment. 2 figs.

  9. Composition and process for the encapsulation and stabilization of radioactive hazardous and mixed wastes

    DOE Patents [OSTI]

    Kalb, Paul D. (21 Barnes Road, Wading River, NY 11792); Colombo, Peter (44 N. Pinelake Dr., Patchogue, NY 11772)

    1997-01-01

    The present invention provides a composition and process for disposal of radioactive, hazardous and mixed wastes. The present invention preferably includes a process for multibarrier encapsulation of radioactive, hazardous and mixed wastes by combining substantially simultaneously dry waste powder, a non-biodegradable thermoplastic polymer and an anhydrous additive in an extruder to form a homogenous molten matrix. The molten matrix may be directed in a "clean" polyethylene liner, allowed to cool, thus forming a monolithic waste form which provides a multibarrier to the dispersion of wastes into the environment.

  10. Composition and process for the encapsulation and stabilization of radioactive, hazardous and mixed wastes

    DOE Patents [OSTI]

    Kalb, P.D.; Colombo, P.

    1999-07-20

    The present invention provides a composition and process for disposal of radioactive, hazardous and mixed wastes. The present invention preferably includes a process for multibarrier encapsulation of radioactive, hazardous and mixed wastes by combining substantially simultaneously dry waste powder, a non-biodegradable thermoplastic polymer and an anhydrous additive in an extruder to form a homogeneous molten matrix. The molten matrix may be directed in a clean'' polyethylene liner, allowed to cool, thus forming a monolithic waste form which provides a multibarrier to the dispersion of wastes into the environment. 2 figs.

  11. Immobilization of sodium nitrate waste with polymers: Topical report

    SciTech Connect (OSTI)

    Franz, E.M.; Heiser, J.H. III; Colombo, P.

    1987-04-01

    This report describes the development of solidification systems for sodium nitrate waste. Sodium nitrate waste was solidified in the polymers polyethylene, polyester-styrene (PES), and water-extendible polyester-styrene (WEP). Evaluations were made of the properties of waste forms containing various amounts of sodium nitrate by leaching immersion in water, measuring compressive strengths and by the EPA Extraction Procedure. Results of the leaching test are presented as cumulative fraction leached (CFL), incremental leaching rate, and average leaching indices (LI). For waste forms containing 30 to 70 wt% sodium nitrate, the CFL ranged from 9.0 x 10/sup -3/ to 7.3 x 10/sup -1/ and the LI from 11 to 7.8. After ninety days immersion in water, the compressive strengths ranged from 720 psi to 2550 psi. The nitrate releases from these samples using the EPA Extraction Procedure were below 500 ppM. The nitrate releases from PES waste forms were similar to those from polyethylene waste forms at the same waste loadings. The compressive yield strengths, measured after ninety-day immersion in water, ranged between 2070 and 7710 psi. In the case of WEP waste forms, only 30 wt% loaded samples passed the immersion test. 23 refs., 24 figs., 12 tabs.

  12. Waste-to-Energy: Waste Management and Energy Production Opportunities...

    Office of Environmental Management (EM)

    Waste-to-Energy: Waste Management and Energy Production Opportunities Waste-to-Energy: Waste Management and Energy Production Opportunities July 24, 2014 9:00AM to 3:30PM EDT U.S....

  13. Electronic waste disassembly with industrial waste heat

    E-Print Network [OSTI]

    2013-01-01

    equipment for automatic dismantling of electronic componentsthe technology acceptance for dismantling of waste printedR. Research on with dismantling of PCB mounted electronic

  14. Bioelectrochemical Integration of Waste Heat Recovery, Waste...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    oxygen demand (COD) and availability of low-grade waste heat sources. The pulp and paper industry and other industries are also potential MHRC users. Project Description This...

  15. Bioelectrochemical Integration of Waste Heat Recovery, Waste...

    Broader source: Energy.gov (indexed) [DOE]

    - Allentown, PA A microbial reverse electrodialysis technology will be combined with waste heat recovery to convert effluents into electricity and chemical products, including...

  16. Properties of vitrified Rocky Flats TRUW with different waste loadings

    SciTech Connect (OSTI)

    Eddy, T.L.; Sears, J.W.; Grandy, J.D.; Miley, D.V.; Erickson, A.W.; Fransworth, R.N.; Larsen, E.D.

    1994-07-01

    One of the major waste streams at the Idaho National Laboratory (INEL) is a combination of the Rocky Flats Plant 1st and 2nd stage sludges (hydrated metal oxides or H-series), which constitutes about 20 wt % of the buried waste. A similar mass fraction is in interim storage. The buried waste is commingled with about five times as much soil that has become contaminated as the containers have deteriorated. The purpose of this paper is to report on waste form property variations of the H-series waste melted with various fractions of soil, plus volatile and hazardous metals and transuranic surrogates. Optimally, the waste form will minimize the bulk leach rate, maximize the volume reduction, minimize the additives needed, and stabilize the transuranic nuclides. Topics to be discussed include the input and final compositions, the melting and crystallization processes, the test results, and conclusions.

  17. Process for removing sulfate anions from waste water

    DOE Patents [OSTI]

    Nilsen, David N. (Lebanon, OR); Galvan, Gloria J. (Albany, OR); Hundley, Gary L. (Corvallis, OR); Wright, John B. (Albany, OR)

    1997-01-01

    A liquid emulsion membrane process for removing sulfate anions from waste water is disclosed. The liquid emulsion membrane process includes the steps of: (a) providing a liquid emulsion formed from an aqueous strip solution and an organic phase that contains an extractant capable of removing sulfate anions from waste water; (b) dispersing the liquid emulsion in globule form into a quantity of waste water containing sulfate anions to allow the organic phase in each globule of the emulsion to extract and absorb sulfate anions from the waste water and (c) separating the emulsion including its organic phase and absorbed sulfate anions from the waste water to provide waste water containing substantially no sulfate anions.

  18. Hanford Site annual dangerous waste report: Volume 1, Part 2, Generator dangerous waste report, dangerous waste

    SciTech Connect (OSTI)

    NONE

    1994-12-31

    This report contains information on hazardous materials at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, waste number, weight, and waste designation.

  19. Report to Congress and the Secretary of Energy: Appendix A 41 U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    extensively on issues related to nuclear waste management and is co- editor of Radioactive Waste FormsReport to Congress and the Secretary of Energy: Appendix A 41 APPENDIX A U.S. NUCLEAR WASTE A 43 Rodney C. Ewing, Ph.D. Chairman Dr. Rodney C. Ewing was appointed to the U.S. Nuclear Waste

  20. Radioactive Waste Management Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09

    This Manual further describes the requirements and establishes specific responsibilities for implementing DOE O 435.1, Radioactive Waste Management, for the management of DOE high-level waste, transuranic waste, low-level waste, and the radioactive component of mixed waste. The purpose of the Manual is to catalog those procedural requirements and existing practices that ensure that all DOE elements and contractors continue to manage DOE's radioactive waste in a manner that is protective of worker and public health and safety, and the environment. Does not cancel other directives.

  1. Testing and Disposal Strategy for Secondary Wastes from Vitrification of Sodium-Bearing Waste at Idaho Nuclear Technology and Engineering Center

    SciTech Connect (OSTI)

    Herbst, Alan K.

    2002-01-02

    The Idaho National Engineering and Environmental Laboratory (INEEL) is considering vitrification to process liquid sodium-bearing waste. Preliminary studies were completed to evaluate the potential secondary wastes comprise acidic and caustic scrubber solutions, HEPA filters, activated carbon, and ion exchange media. Possible treatment methods, waste forms, and disposal sites are evaluated from radiological and mercury contamination estimates.

  2. Testing and Disposal Strategy for Secondary Wastes from Vitrification of Sodium-Bearing Waste at the Idaho Nuclear Technology and Engineering Center

    SciTech Connect (OSTI)

    Herbst, Alan Keith

    2002-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is considering vitrification to process liquid sodium-bearing waste. Preliminary studies were completed to evaluate the potential secondary wastes comprise acidic and caustic scrubber solutions, HEPA filters, activated carbon, and ion exchange media. Possible treatment methods, waste forms, and disposal sites are evaluated from radiological and mercury contamination estimates.

  3. Polymer solidification of mixed wastes at the Rocky Flats Plant

    SciTech Connect (OSTI)

    Faucette, A.M.; Logsdon, B.W.; Lucerna, J.J.; Yudnich, R.J.

    1994-02-01

    The Rocky Flats Plant is pursuing polymer solidification as a viable treatment option for several mixed waste streams that are subject to land disposal restrictions within the Resource Conservation and Recovery Act provisions. Tests completed to date using both surrogate and actual wastes indicate that polyethylene microencapsulation is a viable treatment option for several mixed wastes at the Rocky Flats Plant, including nitrate salts, sludges, and secondary wastes such as ash. Treatability studies conducted on actual salt waste demonstrated that the process is capable of producing waste forms that comply with all applicable regulatory criteria, including the Toxicity Characteristic Leaching Procedure. Tests have also been conducted to evaluate the feasibility of macroencapsulating certain debris wastes in polymers. Several methods and plastics have been tested for macroencapsulation, including post-consumer recycle and regrind polyethylene.

  4. Method for acid oxidation of radioactive, hazardous, and mixed organic waste materials

    DOE Patents [OSTI]

    Pierce, Robert A. (Aiken, SC); Smith, James R. (Corrales, NM); Ramsey, William G. (Aiken, SC); Cicero-Herman, Connie A. (Aiken, SC); Bickford, Dennis F. (Folly Beach, SC)

    1999-01-01

    The present invention is directed to a process for reducing the volume of low level radioactive and mixed waste to enable the waste to be more economically stored in a suitable repository, and for placing the waste into a form suitable for permanent disposal. The invention involves a process for preparing radioactive, hazardous, or mixed waste for storage by contacting the waste starting material containing at least one organic carbon-containing compound and at least one radioactive or hazardous waste component with nitric acid and phosphoric acid simultaneously at a contacting temperature in the range of about 140.degree. C. to about 210 .degree. C. for a period of time sufficient to oxidize at least a portion of the organic carbon-containing compound to gaseous products, thereby producing a residual concentrated waste product containing substantially all of said radioactive or inorganic hazardous waste component; and immobilizing the residual concentrated waste product in a solid phosphate-based ceramic or glass form.

  5. Sorbent Testing for the Solidification of Organic Process Waste streams from the Radiochemical Engineering Development Center at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Bickford, J.; Foote, M. [MSE Technology Applications, Inc., Montana (United States); Taylor, P. [Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States)

    2008-07-01

    The U.S. Department of Energy (DOE) has tasked MSE Technology Applications, Inc. (MSE) with evaluating various sorbents to solidify the radioactive liquid organic waste from the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL). REDC recovers and purifies heavy elements (berkelium, californium, einsteinium, and fermium) from irradiated targets for research and industrial applications. Both aqueous and organic waste streams are discharged from REDC. Organic waste is generated from the plutonium/uranium extraction (PUREX), Cleanex, and Pubex processes.1 The PUREX waste derives from an organic-aqueous isotope separation process for plutonium and uranium fission products, the Cleanex waste derives from the removal of fission products and other impurities from the americium/curium product, and the Pubex waste is derived from the separation process of plutonium from dissolved targets. An aqueous waste stream is also produced from these separation processes. MSE has been tasked to test a grouting formula for the aqueous waste stream that includes specially formulated radioactive shielding materials developed by Science and Technology Applications, LLC. This paper will focus on the sorbent testing work. Based on work performed at Savannah River Site (SRS) (Refs. 1, 2), ORNL tested and evaluated three sorbents capable of solidifying the PUREX, Pubex, and Cleanex waste streams and a composite of the three organic waste streams: Imbiber Beads{sup R} IMB230301 (Imbiber Beads), Nochar A610 Petro Bond, and Petroset II Granular{sup TM} (Petroset II-G). Surrogates of the PUREX, Pubex, Cleanex, and a composite organic waste were used for the bench-scale testing. Recommendations resulting from the ORNL testing included follow-on testing by MSE for two of the three sorbents: Nochar Petro Bond and Petroset II-G. MSE recommended that another clay sorbent, Organoclay BM-QT-199, be added to the test sequence. The sorbent/surrogate combinations were tested at bench scale, 19-liter (L) [5-gallon (gal)] bucket scale, and 208-L (55-gal) drum scale. The testing performed by MSE will help ORNL select the right solidification materials and wasteform generation methods for the design of a new treatment facility. The results could also be used to help demonstrate that ORNL could meet the waste acceptance criteria for the ultimate disposal site for the waste-forms. The organics will be solidified as transuranic waste for disposal at the Waste Isolation Pilot Plant, and the aqueous waste stream will be grouted and disposed of at the Nevada Test Site as low-level waste if real waste testing indicates similar results to the surrogate testing. The objective of this work was to identify a sorbent capable of solidifying PUREX, Pubex, and Cleanex organic wastes individually and a composite of the three organic waste streams. The sorbent and surrogate combinations must also be compatible with processing equipment and maintain stability under a variety of conditions that could occur during storage/shipment of the solidified wastes. (authors)

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

    SciTech Connect (OSTI)

    Deffenbaugh, M.L.

    1997-09-04

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

  7. FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

    SciTech Connect (OSTI)

    Jones, R.; Carter, J.

    2010-10-13

    The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S; (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated; (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass; and (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

  8. FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

    SciTech Connect (OSTI)

    Carter, J.

    2011-01-03

    The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S. (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated. (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass. (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

  9. Copenhagen Waste Management and Incineration

    E-Print Network [OSTI]

    Columbia University

    Copenhagen Waste Management and Incineration Florence, April 24 2009 Julie B. Svendsen 24 20092 Presentation · General introduction to Copenhagen Waste Management System · National incentives · Waste Management plan 2012 · Incineration plants #12;Florence, April 24 20093 Copenhagen Waste

  10. Environmental assessment: special isotope separation process selection

    SciTech Connect (OSTI)

    Not Available

    1986-04-01

    This Environmental Assessment (EA) evaluates the differences in potential environmental impacts between two plutonium Special Isotope Separation (SIS) technologies: Atomic Vapor Laser Isotope Separation (AVLIS) and Molecular Laser Isotope Separation (MLIS). Both SIS technologies use PuO/sub 2/ as feed; AVLIS converts feed to plutonium metal and MLIS converts feed to PuF/sub 6/. The AVLIS process uses laser energy to selectively photoionize and electrostatically separate plutonium isotopes from an atomic vapor stream. The MLIS process uses laser energy to selectively disassociate specific isotopes of plutonium in the form of PuF/sub 6/ molecules to create PuF/sub 5/ for collection and further processing. Both processes produce plutonium metal as their product. An evaluation of differences in potential environmental impacts attributed to the construction of an SIS facility, based on either technology, included a comparison of construction materials, land areas required, and the size of the design and construction workforce. The differences in potential environmental impacts from operating an SIS facility were also compared. No large differences in potential environmental impacts would be expected from the use of process chemicals. An AVLIS or an MLIS facility would produce operating effluents that would meet all applicable radiation, chemical, and hazardous waste standards and would be constructed to protect workers, the public and the environment. This EA has not revealed any significant differences in the potential environmental impacts that could occur as a result of deploying either the AVLIS or the MLIS Special Isotope Separation technology.

  11. Waste Management and WasteWaste Management and Waste--toto--EnergyEnergy Status in SingaporeStatus in Singapore

    E-Print Network [OSTI]

    Columbia University

    ;20031970 The Solid Waste Challenge Waste Explosion 1,200 t/d1,200 t/d 6,900 t/d6,900 t/d #12;Waste ManagementWaste Management and WasteWaste Management and Waste--toto--EnergyEnergy Status in Singapore #12;Singapore's Waste Management · In 2003, 6877 tonnes/day (2.51 M tonnes/year) of MSW collected

  12. Hazardous Waste Management (Delaware)

    Broader source: Energy.gov [DOE]

    The act authorizes the Delaware Department of Natural Resources and Environment Control (DNREC) to regulate hazardous waste and create a program to manage sources of hazardous waste. The act...

  13. Hanford Tank Waste Residuals

    Office of Environmental Management (EM)

    Hanford Tank Waste Residuals DOE HLW Corporate Board November 6, 2008 Chris Kemp, DOE ORP Bill Hewitt, YAHSGS LLC Hanford Tanks & Tank Waste * Single-Shell Tanks (SSTs) - 27...

  14. Nuclear Waste Partnership, LLC

    Office of Environmental Management (EM)

    Nuclear Waste Partnership, LLC Waste Isolation Pilot Plant Report from the Department of Energy Voluntary Protection Program Onsite Review March 17-27, 2015 U.S. Department of...

  15. Pet Waste Management 

    E-Print Network [OSTI]

    Mechell, Justin; Lesikar, Bruce J.

    2008-08-28

    About 1 million pounds of dog waste is deposited daily in North Texas alone. That's why proper disposal of pet waste can make a big difference in the environment. 5 photos, 2 pages...

  16. Vitrification of IFR and MSBR halide salt reprocessing wastes

    SciTech Connect (OSTI)

    Siemer, D.D.

    2013-07-01

    Both of the genuinely sustainable (breeder) nuclear fuel cycles (IFR - Integral Fast Reactor - and MSBR - Molten Salt Breeder Reactor -) studied by the USA's national laboratories would generate high level reprocessing waste (HLRW) streams consisting of a relatively small amount ( about 4 mole %) of fission product halide (chloride or fluoride) salts in a matrix comprised primarily (about 95 mole %) of non radioactive alkali metal halide salts. Because leach resistant glasses cannot accommodate much of any of the halides, most of the treatment scenarios previously envisioned for such HLRW have assumed a monolithic waste form comprised of a synthetic analog of an insoluble crystalline halide mineral. In practice, this translates to making a 'substituted' sodalite ('Ceramic Waste Form') of the IFR's chloride salt-based wastes and fluoroapatite of the MSBR's fluoride salt-based wastes. This paper discusses my experimental studies of an alternative waste management scenario for both fuel cycles that would separate/recycle the waste's halide and immobilize everything else in iron phosphate (Fe-P) glass. It will describe both how the work was done and what its results indicate about how a treatment process for both of those wastes should be implemented (fluoride and chloride behave differently). In either case, this scenario's primary advantages include much higher waste loadings, much lower overall cost, and the generation of a product (glass) that is more consistent with current waste management practices. (author)

  17. Waste vitrification projects throughout the US initiated by SRS

    SciTech Connect (OSTI)

    Jantzen, C.M.; Whitehouse, J.C.; Smith, M.E.; Ramsey, W.G.; Pickett, J.B.

    1996-05-01

    Technologies are being developed by the US Department of Energy (DOE) Nuclear Facility sites to convert high-level, low-level, and mixed wastes to a solid stabilized waste form for permanent disposal. Vitrification is one of the most important and environmentally safest technologies being developed. The Environmental Protection Agency (EPA) has declared vitrification the Best Demonstrated Available Technology (BDAT) for high-level radioactive waste and produced a Handbook of Vitrification Technologies for Treatment of Hazardous and Radioactive Waste. The Defense Waste Processing Facility (DWPF) being tested at Savannah River Site (SRS) will soon begin vitrifying the high-level waste at SRS. The DOE Office of Technology Development (OTD) has taken the position that mixed waste needs to be stabilized to the highest level reasonably possible to ensure that the resulting waste forms will meet both the current and future regulatory specifications. Vitrification produces durable waste forms at volume reductions up to 97%. Large reductions in volume minimize long-term storage costs making vitrification cost effective on a life cycle basis.

  18. Solid waste handling

    SciTech Connect (OSTI)

    Parazin, R.J.

    1995-05-31

    This study presents estimates of the solid radioactive waste quantities that will be generated in the Separations, Low-Level Waste Vitrification and High-Level Waste Vitrification facilities, collectively called the Tank Waste Remediation System Treatment Complex, over the life of these facilities. This study then considers previous estimates from other 200 Area generators and compares alternative methods of handling (segregation, packaging, assaying, shipping, etc.).

  19. Vitrification of organics-containing wastes

    DOE Patents [OSTI]

    Bickford, D.F.

    1997-09-02

    A process is described for stabilizing organics-containing waste materials and recovering metals therefrom, and a waste glass product made according to the process is also disclosed. Vitrification of wastes such as organic ion exchange resins, electronic components and the like can be accomplished by mixing at least one transition metal oxide with the wastes, and, if needed, glass formers to compensate for a shortage of silicates or other glass formers in the wastes. The transition metal oxide increases the rate of oxidation of organic materials in the wastes to improve the composition of the glass-forming mixture: at low temperatures, the oxide catalyzes oxidation of a portion of the organics in the waste; at higher temperatures, the oxide dissolves and the resulting oxygen ions oxidize more of the organics; and at vitrification temperatures, the metal ions conduct oxygen into the melt to oxidize the remaining organics. In addition, the transition metal oxide buffers the redox potential of the glass melt so that metals such as Au, Pt, Ag, and Cu separate from the melt in the metallic state and can be recovered. After the metals are recovered, the remainder of the melt is allowed to cool and may subsequently be disposed of. The product has good leaching resistance and can be disposed of in an ordinary landfill, or, alternatively, used as a filler in materials such as concrete, asphalt, brick and tile. 1 fig.

  20. Hot isostatic press waste option study report

    SciTech Connect (OSTI)

    Russell, N.E.; Taylor, D.D.

    1998-02-01

    A Settlement Agreement between the Department of Energy and the State of Idaho mandates that all high-level radioactive waste now stored at the Idaho Chemical Processing Plant be treated so that it is ready to move out of Idaho for disposal by the target date of 2035. This study investigates the immobilization of all Idaho Chemical Processing Plant calcine, including calcined sodium bearing waste, via the process known as hot isostatic press, which produces compact solid waste forms by means of high temperature and pressure (1,050 C and 20,000 psi), as the treatment method for complying with the settlement agreement. The final waste product would be contained in stainless-steel canisters, the same type used at the Savannah River Site for vitrified waste, and stored at the Idaho National Engineering and Environmental Laboratory until a national geological repository becomes available for its disposal. The waste processing period is from 2013 through 2032, and disposal at the High Level Waste repository will probably begin sometime after 2065.

  1. Secondary Waste Considerations for Vitrification of Sodium-Bearing Waste at the Idaho Nuclear Technology and Engineering Center FY-2001 Status Report

    SciTech Connect (OSTI)

    Herbst, A.K.; Kirkham, R.J.; Losinski, S.J.

    2002-09-26

    The Idaho Nuclear Technology and Engineering Center (INTEC) is considering vitrification to process liquid sodium-bearing waste. Preliminary studies were completed to evaluate the potential secondary wastes from the melter off-gas clean up systems. Projected secondary wastes comprise acidic and caustic scrubber solutions, HEPA filters, activated carbon, and ion exchange media. Possible treatment methods, waste forms, and disposal sites are evaluated from radiological and mercury contamination estimates.

  2. Secondary Waste Considerations for Vitrification of Sodium-Bearing Waste at the Idaho Nuclear Techology and Engineering Center FY-2001 Status Report

    SciTech Connect (OSTI)

    Herbst, Alan Keith; Kirkham, Robert John; Losinski, Sylvester John

    2001-09-01

    The Idaho Nuclear Technology and Engineering Center (INTEC) is considering vitrification to process liquid sodium-bearing waste. Preliminary studies were completed to evaluate the potential secondary wastes from the melter off-gas clean up systems. Projected secondary wastes comprise acidic and caustic scrubber solutions, HEPA filters, activated carbon, and ion exchange media. Possible treatment methods, waste forms, and disposal sites are evaluated from radiological and mercury contamination estimates.

  3. Directory of Energy Data Collection Forms: Forms in use as of October 1996

    SciTech Connect (OSTI)

    1996-12-01

    This is the twentieth edition of the Directory of Energy Collection Data Forms, an authoritative listing of selected public use forms currently used as basic energy information gathering tools by the Department of Energy.

  4. Radioactive tank waste remediation focus area

    SciTech Connect (OSTI)

    1996-08-01

    EM`s Office of Science and Technology has established the Tank Focus Area (TFA) to manage and carry out an integrated national program of technology development for tank waste remediation. The TFA is responsible for the development, testing, evaluation, and deployment of remediation technologies within a system architecture to characterize, retrieve, treat, concentrate, and dispose of radioactive waste stored in the underground stabilize and close the tanks. The goal is to provide safe and cost-effective solutions that are acceptable to both the public and regulators. Within the DOE complex, 335 underground storage tanks have been used to process and store radioactive and chemical mixed waste generated from weapon materials production and manufacturing. Collectively, thes tanks hold over 90 million gallons of high-level and low-level radioactive liquid waste in sludge, saltcake, and as supernate and vapor. Very little has been treated and/or disposed or in final form.

  5. Medical waste treatment and decontamination system

    DOE Patents [OSTI]

    Wicks, George G. (Aiken, SC); Schulz, Rebecca L. (Aiken, SC); Clark, David E. (Gainesville, FL)

    2001-01-01

    The invention discloses a tandem microwave system consisting of a primary chamber in which hybrid microwave energy is used for the controlled combustion of materials. A second chamber is used to further treat the off-gases from the primary chamber by passage through a susceptor matrix subjected to additional hybrid microwave energy. The direct microwave radiation and elevated temperatures provide for significant reductions in the qualitative and quantitative emissions of the treated off gases. The tandem microwave system can be utilized for disinfecting wastes, sterilizing materials, and/or modifying the form of wastes to solidify organic or inorganic materials. The simple design allows on-site treatment of waste by small volume waste generators.

  6. Tandem microwave waste remediation and decontamination system

    DOE Patents [OSTI]

    Wicks, George G. (North Aiken, SC); Clark, David E. (Gainesville, FL); Schulz, Rebecca L. (Gainesville, FL)

    1999-01-01

    The invention discloses a tandem microwave system consisting of a primary chamber in which microwave energy is used for the controlled combustion of materials. A second chamber is used to further treat the off-gases from the primary chamber by passage through a susceptor matrix subjected to additional microwave energy. The direct microwave radiation and elevated temperatures provide for significant reductions in the qualitative and quantitative emissions of the treated off gases. The tandem microwave system can be utilized for disinfecting wastes, sterilizing materials, and/or modifying the form of wastes to solidify organic or inorganic materials. The simple design allows on-site treatment of waste by small volume waste generators.

  7. High-Level Waste Melter Study Report

    SciTech Connect (OSTI)

    Perez, Joseph M.; Bickford, Dennis F.; Day, Delbert E.; Kim, Dong-Sang; Lambert, Steven L.; Marra, Sharon L.; Peeler, David K.; Strachan, Denis M.; Triplett, Mark B.; Vienna, John D.; Wittman, Richard S.

    2001-07-13

    At the Hanford Site in Richland, Washington, the path to site cleanup involves vitrification of the majority of the wastes that currently reside in large underground tanks. A Joule-heated glass melter is the equipment of choice for vitrifying the high-level fraction of these wastes. Even though this technology has general national and international acceptance, opportunities may exist to improve or change the technology to reduce the enormous cost of accomplishing the mission of site cleanup. Consequently, the U.S. Department of Energy requested the staff of the Tanks Focus Area to review immobilization technologies, waste forms, and modifications to requirements for solidification of the high-level waste fraction at Hanford to determine what aspects could affect cost reductions with reasonable long-term risk. The results of this study are summarized in this report.

  8. FEASIBILITY AND EXPEDIENCE TO VITRIFY NPP OPERATIONAL WASTE

    SciTech Connect (OSTI)

    LIFANOV, F.A.; OJOVAN, M.I.; STEFANOVSKY, S.V.; BURCL, R.

    2003-02-27

    Operational radioactive waste is generated during routine operation of NPP. Process waste is mainly generated by treatment of water from reactor or ancillaries including spent fuel storage pools and some decontamination operations. Typical process wastes of pressurized water reactors (PWR or WWER) are borated water concentrates, whereas typical process wastes of boiling and RBMK type reactors are water concentrates with no boron content. NPP operational wastes are classified as low and intermediate level waste (LILW). NPP operational waste must be solidified in order to ensure safe conditions of storage and disposal. Currently the most promising solidification method for this waste is the vitrification technology. Vitrification of NPP operational waste is a relative new option being developed for last years. Nevertheless there is already accumulated operational experience on vitrifying low and intermediate level waste in Russian Federation at Moscow SIA ''Radon'' vitrification plant. This plant uses the most advanced type induction high frequency melters that facilitate the melting process and significantly reduce the generation of secondary waste and henceforth the overall cost. The plant was put into operation by the end of 1999. It has three operating cold crucible melters with the overall capacity up to 75 kg/h. The vitrification technology comprises a few stages, starting with evaporation of excess water from liquid radioactive waste, followed by batch preparation, glass melting, and ending with vitrified waste blocks and some relative small amounts of secondary waste. First of all since the original waste contain as main component water, this water is removed from waste through evaporation. Then the remaining salt concentrate is mixed with necessary technological additives, thus a glass-forming batch is formed. The batch is fed into melters where the glass melting occurs. From here there are two streams: one is the glass melt containing the most part of radioactivity and second is the off gas flow, which contains off gaseous and aerosol airborne. The melt glass is fed into containers, which are slowly cooled in an annealing tunnel furnace to avoid accumulation of mechanical stresses in the glass. Containers with glass are the final processing product containing the overwhelming part of waste contaminants. The second stream from melter is directed to gas purification system, which is a rather complex system taking into account the necessity to remove from off gas not only radionuclides but also the chemical contaminants. Operation of this purification system leads to generation of a small amount of secondary waste. This waste stream slightly contaminated with volatilized radionuclides is recycled in the same technological scheme. As a result only non-radioactive materials are produced. They are either discharged into environment or reused. Based on the experience gained during operation of vitrification plant one can conclude on high efficiency achieved through vitrification method. Another significant argument on vitrifying NPP operational waste is the minimal impact of vitrified radioactive waste onto environment. Solidified waste shall be disposed of into a near surface disposal facility. Waste forms disposed of in a near-surface wet repository eventually come into contact with groundwater. Engineered structures used or designed to prevent or postpone such contact and the subsequent radionuclide release are complex and often too expensive. Vitrification technologies provide waste forms with excellent resistance to corrosion and gave the basic possibility of maximal simplification of engineered barrier systems. The most simple disposal option is to locate the vitrified waste form packages directly into earthen trenches provided the host rock has the necessary sorption and confinement properties. Such an approach will significantly make simpler the disposal facilities thus contributing both to enhancing safety and economic al efficiency.

  9. Mineralogical textural and compositional data on the alteration of basaltic glass from Kilauea, Hawaii to 300 degrees C: Insights to the corrosion of a borosilicate glass waste-form. [Yucca Mountain Project

    SciTech Connect (OSTI)

    Smith, D.K.

    1990-01-01

    Mineralogical, textural and compositional data accompanying greenschist facies metamorphism (to 300{degrees}C) of basalts of the East Rift Zone (ERZ), Kilauea, Hawaii may be evaluated relative to published and experimental results for the surface corrosion of borosilicate glass. The ERZ alteration sequence is dominated by intermittent palagonite, interlayered smectite-chlorite, chlorite, and actinolite-epidote-anhydrite. Alteration is best developed in fractures and vesicles where surface reaction layers root on the glass matrix forming rinds in excess of 100 microns thick. Fractures control fluid circulation and the alteration sequence. Proximal to the glass surface, palagonite, Fe-Ti oxides and clays replace fresh glass as the surface reaction layer migrates inwards; away from the surface, amphibole, anhydrite, quartz and calcite crystallize from hydrothermal fluids in contact with the glass. The texture and composition of basaltic glass surfaces are similar to those of a SRL-165 glass leached statically for sixty days at 150 {degrees}C. While the ERZ reservoir is a complex open system, conservative comparisons between the alteration of ERZ and synthetic borosilicate glass are warranted. 31 refs., 2 figs.

  10. Management of immunization solid wastes in Kano State, Nigeria

    SciTech Connect (OSTI)

    Oke, I.A. [Civil Engineering Department, Obafemi Awolowo University, Ile-Ife (Nigeria)], E-mail: okeia@oauife.edu.ng

    2008-12-15

    Inadequate management of waste generated from injection activities can have a negative impact on the community and environment. In this paper, a report on immunization wastes management in Kano State (Nigeria) is presented. Eight local governments were selected randomly and surveyed by the author. Solid wastes generated during the Expanded Programme on Immunization were characterised using two different methods: one by weighing the waste and the other by estimating the volume. Empirical data was obtained on immunization waste generation, segregation, storage, collection, transportation, and disposal; and waste management practices were assessed. The study revealed that immunization offices were accommodated in either in local government buildings, primary health centres or community health care centres. All of the stations demonstrated a high priority for segregation of the infectious wastes. It can be deduced from the data obtained that infectious waste ranged from 67.6% to 76.7% with an average of 70.1% by weight, and 36.0% to 46.1% with an average of 40.1% by volume. Non-infectious waste generated ranged from 23.3% to 32.5% with an average of 29.9% by weight and 53.9% to 64.0% with an average of 59.9% by volume. Out of non-infectious waste (NIFW) and infectious waste (IFW), 66.3% and 62.4% by weight were combustible and 33.7% and 37.6% were non-combustible respectively. An assessment of the treatment revealed that open pit burning and burial and small scale incineration were the common methods of disposal for immunization waste, and some immunization centres employed the services of the state or local government owned solid waste disposal board for final collection and disposal of their immunization waste at government approved sites.

  11. Sodium-bearing Waste Treatment Technology Evaluation Report

    SciTech Connect (OSTI)

    Charles M. Barnes; Arlin L. Olson; Dean D. Taylor

    2004-05-01

    Sodium-bearing waste (SBW) disposition is one of the U.S. Department of Energy (DOE) Idaho Operation Office’s (NE-ID) and State of Idaho’s top priorities at the Idaho National Engineering and Environmental Laboratory (INEEL). The INEEL has been working over the past several years to identify a treatment technology that meets NE-ID and regulatory treatment requirements, including consideration of stakeholder input. Many studies, including the High-Level Waste and Facilities Disposition Environmental Impact Statement (EIS), have resulted in the identification of five treatment alternatives that form a short list of perhaps the most appropriate technologies for the DOE to select from. The alternatives are (a) calcination with maximum achievable control technology (MACT) upgrade, (b) steam reforming, (c) cesium ion exchange (CsIX) with immobilization, (d) direct evaporation, and (e) vitrification. Each alternative has undergone some degree of applied technical development and preliminary process design over the past four years. This report presents a summary of the applied technology and process design activities performed through February 2004. The SBW issue and the five alternatives are described in Sections 2 and 3, respectively. Details of preliminary process design activities for three of the alternatives (steam reforming, CsIX, and direct evaporation) are presented in three appendices. A recent feasibility study provides the details for calcination. There have been no recent activities performed with regard to vitrification; that section summarizes and references previous work.

  12. Savannah River Site Marks Waste Processing Milestone with Melter’s 2,000th Waste Canister

    Broader source: Energy.gov [DOE]

    AIKEN, S.C. – The second melter to operate in the 16-year history of the nation’s largest radioactive waste glassification plant shows no signs of slowing after recently pouring its 2,000 canister of glass-formed hazardous waste.

  13. Radioactive Waste Management Basis

    SciTech Connect (OSTI)

    Perkins, B K

    2009-06-03

    The purpose of this Radioactive Waste Management Basis is to describe the systematic approach for planning, executing, and evaluating the management of radioactive waste at LLNL. The implementation of this document will ensure that waste management activities at LLNL are conducted in compliance with the requirements of DOE Order 435.1, Radioactive Waste Management, and the Implementation Guide for DOE Manual 435.1-1, Radioactive Waste Management Manual. Technical justification is provided where methods for meeting the requirements of DOE Order 435.1 deviate from the DOE Manual 435.1-1 and Implementation Guide.

  14. Municipal waste processing apparatus

    DOE Patents [OSTI]

    Mayberry, J.L.

    1988-04-13

    This invention relates to apparatus for processing municipal waste, and more particularly to vibrating mesh screen conveyor systems for removing grit, glass, and other noncombustible materials from dry municipal waste. Municipal waste must be properly processed and disposed of so that it does not create health risks to the community. Generally, municipal waste, which may be collected in garbage trucks, dumpsters, or the like, is deposited in processing areas such as landfills. Land and environmental controls imposed on landfill operators by governmental bodies have increased in recent years, however, making landfill disposal of solid waste materials more expensive. 6 figs.

  15. Radioactive Waste Management Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09

    This Manual further describes the requirements and establishes specific responsibilities for implementing DOE O 435.1, Radioactive Waste Management, for the management of DOE high-level waste, transuranic waste, low-level waste, and the radioactive component of mixed waste. Change 1 dated 6/19/01 removes the requirement that Headquarters is to be notified and the Office of Environment, Safety and Health consulted for exemptions for use of non-DOE treatment facilities. Certified 1-9-07. Admin Chg 2, dated 6-8-11, supersedes DOE M 435.1-1 Chg 1.

  16. Impact assessment of draft DOE Order 5820.2B. Radioactive Waste Technical Support Program

    SciTech Connect (OSTI)

    NONE

    1995-04-01

    The Department of Energy (DOE) has prepared a revision to DOE Order 5820.2A, entitled ``Radioactive Waste Management.`` DOE issued DOE Order 5820.2A in September 1988 and, as the title implies, it covered only radioactive waste forms. The proposed draft order, entitled ``Waste Management,`` addresses the management of both radioactive and nonradioactive waste forms. It also includes spent nuclear fuel, which DOE does not consider a waste. Waste forms covered include hazardous waste, high-level waste, transuranic (TRU) waste, low-level radioactive waste, uranium and thorium mill tailings, mixed waste, and sanitary waste. The Radioactive Waste Technical Support Program (TSP) of Leached Idaho Technologies Company (LITCO) is facilitating the revision of this order. The EM Regulatory Compliance Division (EM-331) has requested that TSP estimate the impacts and costs of compliance with the revised order. TSP requested Dames & Moore to aid in this assessment by comparing requirements in Draft Order 5820.2B to ones in DOE Order 5820.2A and other DOE orders and Federal regulations. The assessment started with a draft version of 5820.2B dated January 14, 1994. DOE has released three updated versions of the draft order since then (dated May 20, 1994; August 26, 1994; and January 23, 1995). Each time DOE revised the order, Dames and Moore updated the assessment work to reflect the text changes. This report reflects the January 23, 1995 version of the draft order.

  17. Hanford site transuranic waste sampling plan

    SciTech Connect (OSTI)

    GREAGER, T.M.

    1999-05-13

    This sampling plan (SP) describes the selection of containers for sampling of homogeneous solids and soil/gravel and for visual examination of transuranic and mixed transuranic (collectively referred to as TRU) waste generated at the U.S. Department of Energy (DOE) Hanford Site. The activities described in this SP will be conducted under the Hanford Site TRU Waste Certification Program. This SP is designed to meet the requirements of the Transuranic Waste Characterization Quality Assurance Program Plan (CAO-94-1010) (DOE 1996a) (QAPP), site-specific implementation of which is described in the Hanford Site Transuranic Waste Characterization Program Quality Assurance Project Plan (HNF-2599) (Hanford 1998b) (QAPP). The QAPP defines the quality assurance (QA) requirements and protocols for TRU waste characterization activities at the Hanford Site. In addition, the QAPP identifies responsible organizations, describes required program activities, outlines sampling and analysis strategies, and identifies procedures for characterization activities. The QAPP identifies specific requirements for TRU waste sampling plans. Table 1-1 presents these requirements and indicates sections in this SP where these requirements are addressed.

  18. Enterprise Assessments Review of the Hanford Site Waste Treatment...

    Energy Savers [EERE]

    by ORP. The mission of the WTP is to process and immobilize the Hanford high-level tank waste into a stable glass form suitable for permanent 2 disposal. The LAW facility is...

  19. Mixed waste: Proceedings

    SciTech Connect (OSTI)

    Moghissi, A.A.; Blauvelt, R.K.; Benda, G.A.; Rothermich, N.E.

    1993-12-31

    This volume contains the peer-reviewed and edited versions of papers submitted for presentation a the Second International Mixed Waste Symposium. Following the tradition of the First International Mixed Waste Symposium, these proceedings were prepared in advance of the meeting for distribution to participants. The symposium was organized by the Mixed Waste Committee of the American Society of Mechanical Engineers. The topics discussed at the symposium include: stabilization technologies, alternative treatment technologies, regulatory issues, vitrification technologies, characterization of wastes, thermal technologies, laboratory and analytical issues, waste storage and disposal, organic treatment technologies, waste minimization, packaging and transportation, treatment of mercury contaminated wastes and bioprocessing, and environmental restoration. Individual abstracts are catalogued separately for the data base.

  20. Ferrocyanide Safety Project: Comparison of actual and simulated ferrocyanide waste properties

    SciTech Connect (OSTI)

    Scheele, R.D.; Burger, L.L.; Sell, R.L.; Bredt, P.R.; Barrington, R.J.

    1994-09-01

    In the 1950s, additional high-level radioactive waste storage capacity was needed to accommodate the wastes that would result from the production of recovery of additional nuclear defense materials. To provide this additional waste storage capacity, the Hanford Site operating contractor developed a process to decontaminate aqueous wastes by precipitating radiocesium as an alkali nickel ferrocyanide; this process allowed disposal of the aqueous waste. The radiocesium scavenging process as developed was used to decontaminate (1) first-cycle bismuth phosphate (BiPO{sub 4}) wastes, (2) acidic wastes resulting from uranium recovery operations, and (3) the supernate from neutralized uranium recovery wastes. The radiocesium scavenging process was often coupled with other scavenging processes to remove radiostrontium and radiocobalt. Because all defense materials recovery processes used nitric acid solutions, all of the wastes contained nitrate, which is a strong oxidizer. The variety of wastes treated, and the occasional coupling of radiostrontium and radiocobalt scavenging processes with the radiocesium scavenging process, resulted in ferrocyanide-bearing wastes having many different compositions. In this report, we compare selected physical, chemical, and radiochemical properties measured for Tanks C-109 and C-112 wastes and selected physical and chemical properties of simulated ferrocyanide wastes to assess the representativeness of stimulants prepared by WHC.