Sample records for liquid waste process

  1. Process for preparing liquid wastes

    DOE Patents [OSTI]

    Oden, Laurance L. (Albany, OR); Turner, Paul C. (Albany, OR); O'Connor, William K. (Lebanon, OR); Hansen, Jeffrey S. (Corvallis, OR)

    1997-01-01T23:59:59.000Z

    A process for preparing radioactive and other hazardous liquid wastes for treatment by the method of vitrification or melting is provided for.

  2. Used powdex resin for liquid waste processing

    SciTech Connect (OSTI)

    Pearson, C.J.; Bramblett, J.W.

    1995-05-01T23:59:59.000Z

    Powdex resin has traditionally been used on the Secondary side for condensate polishing. The resins on these polishers are backwashed periodically based on chemical breakthrough or high differential pressure. Upon the backwash, the ion exchange capacity on the powdex resin is not completely exhausted. In the past, this partially used powdex resin was sluiced to a liner and treated as waste for disposal. In an effort to reduce radioactivity being released from segregated, high chemical concentration wastes, the idea of trying used powdex resin was initiated. In 1992, Duke Power Company began processing problem liquid waste streams with used Secondary powdex and subsequent decanting of the waste water for release. the results have shown significant reductions in the activity of this water. This paper will detail the history, method, and the results of using Secondary powdex for liquid radwaste processing. It will also describe the benefits, such as: (1) Processing waste streams not suitable for bead resin demineralizers. (2) Ability to process large volumes of waste water in a short period of time. (3) Recycling media thought to be useless. (4) > 80% Reduction in activity of water processed. (5) Overall curies reduction. (6) Improved bead demineralizer performance.

  3. Newly Generated Liquid Waste Processing Alternatives Study, Volume 1

    SciTech Connect (OSTI)

    Landman, William Henry; Bates, Steven Odum; Bonnema, Bruce Edward; Palmer, Stanley Leland; Podgorney, Anna Kristine; Walsh, Stephanie

    2002-09-01T23:59:59.000Z

    This report identifies and evaluates three options for treating newly generated liquid waste at the Idaho Nuclear Technology and Engineering Center of the Idaho National Engineering and Environmental Laboratory. The three options are: (a) treat the waste using processing facilities designed for treating sodium-bearing waste, (b) treat the waste using subcontractor-supplied mobile systems, or (c) treat the waste using a special facility designed and constructed for that purpose. In studying these options, engineers concluded that the best approach is to store the newly generated liquid waste until a sodium-bearing waste treatment facility is available and then to co-process the stored inventory of the newly generated waste with the sodium-bearing waste. After the sodium-bearing waste facility completes its mission, two paths are available. The newly generated liquid waste could be treated using the subcontractor-supplied system or the sodium-bearing waste facility or a portion of it. The final decision depends on the design of the sodium-bearing waste treatment facility, which will be completed in coming years.

  4. Transfer Lines to Connect Liquid Waste Facilities and Salt Waste...

    Office of Environmental Management (EM)

    Transfer Lines to Connect Liquid Waste Facilities and Salt Waste Processing Facility Transfer Lines to Connect Liquid Waste Facilities and Salt Waste Processing Facility October...

  5. An evaluation of neutralization for processing sodium-bearing liquid waste

    SciTech Connect (OSTI)

    Chipman, N.A.; Engelgau, G.O.; Berreth, J.R.

    1989-01-01T23:59:59.000Z

    This report addresses an alternative concept for potentially managing the sodium-bearing liquid waste generated at the Idaho Chemical Processing Plant from the current method of calcining a blend of sodium waste and high-level liquid waste. The concept is based on removing the radioactive components from sodium-bearing waste by neutralization and grouting the resulting low-level waste for on-site near-surface disposal. Solidifying the sodium waste as a remote-handled transuranic waste is not considered to be practical because of excessive costs and inability to dispose of the waste in a timely fashion. Although neutralization can remove most radioactive components to provide feed for a solidified low-level waste, and can reduce liquid inventories four to nine years more rapidly than the current practice of blending sodium-bearing liquid waste with first-cycle raffinite, the alternative will require major new facilities and will generate large volumes of low-level waste. Additional facility and operating costs are estimated to be at least $500 million above the current practice of blending and calcining. On-site, low-level waste disposal may be technically difficult and conflict which national and state policies. Therefore, it is recommended that the current practice of calcining a blend of sodium-bearing liquid waste and high-level liquid waste be continued to minimize overall cost and process complexities. 17 refs., 4 figs., 16 tabs.

  6. Process for immobilizing radioactive boric acid liquid wastes

    DOE Patents [OSTI]

    Greenhalgh, Wilbur O. (Richland, WA)

    1986-01-01T23:59:59.000Z

    A method of immobilizing boric acid liquid wastes containing radionuclides by neutralizing the solution and evaporating the resulting precipitate to near dryness. The dry residue is then fused into a reduced volume, insoluble, inert, solid form containing substantially all the radionuclides.

  7. Process for immobilizing radioactive boric acid liquid wastes

    DOE Patents [OSTI]

    Greenhalgh, W.O.

    1984-05-10T23:59:59.000Z

    Disclosed is a method of immobilizing boric acid liquid wastes containing radionuclides by neutralizing the solution and evaporating the resulting precipitate to near dryness. The dry residue is then fused into a reduced volume, insoluble, inert, solid form containing substantially all the radionuclides.

  8. Proceedings of waste stream minimization and utilization innovative concepts: An experimental technology exchange. Volume 2, Industrial liquid waste processing, industrial gaseous waste processing

    SciTech Connect (OSTI)

    Lee, V.E. [ed.; Watts, R.L.

    1993-04-01T23:59:59.000Z

    This two-volume proceedings summarize the results of fifteen innovations that were funded through the US Department of Energy`s Innovative Concept Program. The fifteen innovations were presented at the sixth Innovative Concepts Fair, held in Austin, Texas, on April 22--23, 1993. The concepts in this year`s fair address innovations that can substantially reduce or use waste streams. Each paper describes the need for the proposed concept, the concept being proposed, and the concept`s economics and market potential, key experimental results, and future development needs. The papers are divided into two volumes: Volume 1 addresses innovations for industrial solid waste processing and municipal waste reduction/recycling, and Volume 2 addresses industrial liquid waste processing and industrial gaseous waste processing. Individual reports are indexed separately.

  9. Treatment of liquid radioactive waste using combination of chemical processes with ultrafiltration

    SciTech Connect (OSTI)

    Zabrodsky, V.N.; Davidov, Y.P.; Toropov, I.G.; Glushko, A.S. [Academy of Sciences of Belarus, Minsk (Belarus). Inst. of Radioecological Problems; Efremenkov, V.M. [State Committee on Supervision of Industrial and Nuclear Safety, Minsk (Belarus)

    1993-12-31T23:59:59.000Z

    A combination of chemical processes and ultrafiltration was used for treatment of liquid radioactive wastes containing Cs, Sr, Pu by using large amounts of surfactants and complexing substances. The nature of carriers and conditions of complete separation of Cs and Sr are reported. The mechanisms of the processes are discussed.

  10. MUSHROOM WASTE MANAGEMENT PROJECT LIQUID WASTE MANAGEMENT

    E-Print Network [OSTI]

    of solid and liquid wastes generated at mushroom producing facilities. Environmental guidelines#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

  11. Use of tangential filtration unit for processing liquid waste from nuclear laundries

    SciTech Connect (OSTI)

    Augustin, X.; Buzonniere, A. de [Technicatome, Gif-sur-Yvette (France); Barnier, H. [CEA Cadarache, St. Paul-lez-Durance (France)

    1993-12-31T23:59:59.000Z

    Nuclear facilities produce large quantities of weakly contaminated effluents charged with insoluble and soluble products. In collaboration with CEA, TECHNICATOME has developed an ultrafiltration process for liquid waste from nuclear facilities associated with prior insolubilization of radiochemical activity. This process, seeded ultrafiltration, is based on the use of a decloggable mineral filter media and combines very high separation efficiency with long membrane life. The efficiency of the tangential filtration unit which has been processing effluents from the Cadarache Nuclear Research Center (CEA-France) nuclear facility since mid-1988, has been confirmed on several sites.

  12. Processing liquid radioactive waste by centrifuge and indrum dehydration facility at NPP Philippsburg

    SciTech Connect (OSTI)

    Grundke, E.; Blaser, W. [NPP Philippsburg (Germany)

    1993-12-31T23:59:59.000Z

    Until 1989 the evaporator and filter concentrates have been treated by concreting. The centrifuge facility is used for the liquid waste from laundry, showers and also for processing filter concentrates and evaporator feedwater. The hot high pressure compacting of filter concentrates gives a volume reduction by a factor of 6. The evaporator concentrate is drained in a 200 l drum and this drum is heated by an external heating device. The indrum-dehydration facility reduces the treated volume by a factor of 12 compared with the former cementation.

  13. EXPLORING ENGINEERING CONTROL THROUGH PROCESS MANIPULATION OF RADIOACTIVE LIQUID WASTE TANK CHEMICAL CLEANING

    SciTech Connect (OSTI)

    Brown, A.

    2014-04-27T23:59:59.000Z

    One method of remediating legacy liquid radioactive waste produced during the cold war, is aggressive in-tank chemical cleaning. Chemical cleaning has successfully reduced the curie content of residual waste heels in large underground storage tanks; however this process generates significant chemical hazards. Mercury is often the bounding hazard due to its extensive use in the separations process that produced the waste. This paper explores how variations in controllable process factors, tank level and temperature, may be manipulated to reduce the hazard potential related to mercury vapor generation. When compared using a multivariate regression analysis, findings indicated that there was a significant relationship between both tank level (p value of 1.65x10{sup -23}) and temperature (p value of 6.39x10{sup -6}) to the mercury vapor concentration in the tank ventilation system. Tank temperature showed the most promise as a controllable parameter for future tank cleaning endeavors. Despite statistically significant relationships, there may not be confidence in the ability to control accident scenarios to below mercury’s IDLH or PAC-III levels for future cleaning initiatives.

  14. Biohazardous Waste Disposal Guidelines Sharps Waste Solid Lab Waste Liquid Waste Animals Pathological Waste

    E-Print Network [OSTI]

    Tsien, Roger Y.

    Biohazardous Waste Disposal Guidelines Sharps Waste Solid Lab Waste Liquid Waste Animals Pathological Waste Description Biohazard symbol Address: UCSD 9500 Gilman Drive La Jolla, CA 92093 (858) 534) and identity of liquid waste Biohazard symbol Address: UCSD 9500 Gilman Drive La Jolla, CA 92093 (858) 534

  15. Biohazardous Waste Disposal Guidelines Sharps Waste Solid Lab Waste Liquid Waste Animals Pathological Waste

    E-Print Network [OSTI]

    Tsien, Roger Y.

    2/2009 Biohazardous Waste Disposal Guidelines Sharps Waste Solid Lab Waste Liquid Waste Animals Pathological Waste Description Biohazard symbol Address: UCSD 200 West Arbor Dr. San Diego, CA 92103 (619 (9:1) OR Biohazard symbol (if untreated) and identity of liquid waste Biohazard symbol Address

  16. Waste processing air cleaning

    SciTech Connect (OSTI)

    Kriskovich, J.R.

    1998-07-27T23:59:59.000Z

    Waste processing and preparing waste to support waste processing relies heavily on ventilation. Ventilation is used at the Hanford Site on the waste storage tanks to provide confinement, cooling, and removal of flammable gases.

  17. Salt Waste Processing Initiatives

    Office of Environmental Management (EM)

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

  18. Bubblers Speed Nuclear Waste Processing at SRS

    SciTech Connect (OSTI)

    None

    2010-11-14T23:59:59.000Z

    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.

  19. Bubblers Speed Nuclear Waste Processing at SRS

    ScienceCinema (OSTI)

    None

    2014-08-06T23:59:59.000Z

    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.

  20. Method for treating liquid wastes

    DOE Patents [OSTI]

    Katti, K.V.; Volkert, W.A.; Singh, P.; Ketring, A.R.

    1995-12-26T23:59:59.000Z

    The method of treating liquid waste in a media is accomplished by exposing the media to phosphinimines and sequestering {sup 99}Tc from the media by the phosphinimine (PN) functionalities. The system for treating the liquid waste in the media includes extraction of {sup 99}TcO{sub 4}{sup {minus}} from aqueous solutions into organic solvents or mixed organic/polar media, extraction of {sup 99}Tc from solutions on a solid matrix by using a container containing PN functionalities on solid matrices including an inlet and outlet for allowing flow of media through an immobilized phosphinimine ligand system contained within the container. Also, insoluble suspensions of phosphinimine functionalities on solid matrices in liquid solutions or present on supported liquid membranes (SLM) can be used to sequester {sup 99}Tc from those liquids. 6 figs.

  1. Method for treating liquid wastes

    DOE Patents [OSTI]

    Katti, Kattesh V. (Columbia, MO); Volkert, Wynn A. (Columbia, MO); Singh, Prahlad (Columbia, MO); Ketring, Alan R. (Columbia, MO)

    1995-01-01T23:59:59.000Z

    The method of treating liquid waste in a media is accomplished by exposing the media to phosphinimines and sequestering .sup.99 Tc from the media by the phosphinimine (PN) functionalities. The system for treating the liquid waste in the media includes extraction of .sup.99 TcO.sub.4.sup.- from aqueous solutions into organic solvents or mixed organic/polar media, extraction of .sup.99 Tc from solutions on a solid matrix by using a container containing PN functionalities on solid matrices including an inlet and outlet for allowing flow of media through an immobilized phosphinimine ligand system contained within the container. Also, insoluble suspensions of phosphinimine functionalities on solid matrices in liquid solutions or present on supported liquid membranes (SLM) can be used to sequester .sup.99 Tc from those liquids.

  2. Development And Initial Testing Of Off-Gas Recycle Liquid From The WTP Low Activity Waste Vitrification Process - 14333

    SciTech Connect (OSTI)

    McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.; Taylor-Pashow, Kathryn M.; Adamson, Duane J.; Crawford, Charles L.; Morse, Megan M.

    2014-01-07T23:59:59.000Z

    The Waste Treatment and Immobilization Plant (WTP) process flow was designed to pre-treat feed from the Hanford tank farms, separate it into a High Level Waste (HLW) and Low Activity Waste (LAW) fraction and vitrify each fraction in separate facilities. Vitrification of the waste generates an aqueous condensate stream from the off-gas processes. This stream originates from two off-gas treatment unit operations, the Submerged Bed Scrubber (SBS) and the Wet Electrospray Precipitator (WESP). Currently, the baseline plan for disposition of the stream from the LAW melter is to recycle it to the Pretreatment facility where it gets evaporated and processed into the LAW melter again. If the Pretreatment facility is not available, the baseline disposition pathway is not viable. Additionally, some components in the stream are volatile at melter temperatures, thereby accumulating to high concentrations in the scrubbed stream. It would be highly beneficial to divert this stream to an alternate disposition path to alleviate the close-coupled operation of the LAW vitrification and Pretreatment facilities, and to improve long-term throughput and efficiency of the WTP system. In order to determine an alternate disposition path for the LAW SBS/WESP Recycle stream, a range of options are being studied. A simulant of the LAW Off-Gas Condensate was developed, based on the projected composition of this stream, and comparison with pilot-scale testing. The primary radionuclide that vaporizes and accumulates in the stream is Tc-99, but small amounts of several other radionuclides are also projected to be present in this stream. The processes being investigated for managing this stream includes evaporation and radionuclide removal via precipitation and adsorption. During evaporation, it is of interest to investigate the formation of insoluble solids to avoid scaling and plugging of equipment. Key parameters for radionuclide removal include identifying effective precipitation or ion adsorption chemicals, solid-liquid separation methods, and achievable decontamination factors. Results of the radionuclide removal testing indicate that the radionuclides, including Tc-99, can be removed with inorganic sorbents and precipitating agents. Evaporation test results indicate that the simulant can be evaporated to fairly high concentration prior to formation of appreciable solids, but corrosion has not yet been examined.

  3. Development and results of experimental testing of electromembrane process for liquid active waste purification

    SciTech Connect (OSTI)

    Martinov, B.V.; Smirnov, V.V.; Tugolukov, B.B.; Belyakov, Y.A. [A.A. Bochvar All Russian Scientific Research, Moscow (Russian Federation). Inst. of Inorganic Materials

    1993-12-31T23:59:59.000Z

    This paper discusses the results of studies on electromembrane purification. The concentration of salts in active wastes arising from decontamination is more than 3--5 g/l. For these investigations a solution was chosen that had arisen from the decontamination of metallic items by a two-bath method using permanganate-alkali in the first stage and nitrogen oxalic acid in the second stage. The total salt content of mixed acid and alkaline solutions was 3.0 g/l, with a pH of 8.5 and total beta-activity of 6 {times} 10{sup {minus}8} Ci/l.

  4. Liquid-Liquid Extraction Processes

    E-Print Network [OSTI]

    Fair, J. R.; Humphrey, J. L.

    1983-01-01T23:59:59.000Z

    Liquid-liquid extraction is the separation of one or more components of a liquid solution by contact with a second immiscible liquid called the solvent. If the components in the original liquid solution distribute themselves differently between...

  5. Municipal waste processing apparatus

    DOE Patents [OSTI]

    Mayberry, J.L.

    1988-04-13T23:59:59.000Z

    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.

  6. Liquid and Gaseous Waste Operations Department Annual Operating Report, CY 1993

    SciTech Connect (OSTI)

    Maddox, J.J.; Scott, C.B.

    1994-02-01T23:59:59.000Z

    This report summarizes the activities of the waste management operations section of the liquid and gaseous waste operations department at ORNL for 1993. The process waste, liquid low-level waste, gaseous waste systems activities are reported, as well as the low-level waste solidification project. Upgrade activities is the various waste processing and treatment systems are summarized. A maintenance activity overview is provided, and program management, training, and other miscellaneous activities are covered.

  7. Unreviewed Safety Question Determination - Processing Waste in...

    Office of Environmental Management (EM)

    Unreviewed Safety Question Determination - Processing Waste in the Waste Characterization Glovebox Unreviewed Safety Question Determination - Processing Waste in the Waste...

  8. MEASUREMENTS TAKEN IN SUPPORT OF QUALIFICATION OF PROCESSING SAVANNAH RIVER SITE LOW-LEVEL LIQUID WASTE INTO SALTSTONE

    SciTech Connect (OSTI)

    Reigel, M.; Bibler, N.; Diprete, C.; Cozzi, A.; Staub, A.; Ray, J.

    2010-01-27T23:59:59.000Z

    The Saltstone Facility at the Savannah River Site (SRS) immobilizes low-level liquid waste into Saltstone to be disposed of in the Z-Area Saltstone Disposal Facility, Class Three Landfill. In order to meet the permit conditions and regulatory limits set by the South Carolina Department of Health and Environmental Control (SCDHEC), the Resource Conservation and Recovery Act (RCRA) and the Environmental Protection Agency (EPA), both the low-level salt solution and Saltstone samples are analyzed quarterly. Waste acceptance criteria (WAC) are designed to confirm the salt solution sample from the Tank Farm meets specific radioactive and chemical limits. The toxic characteristic leaching procedure (TCLP) is used to confirm that the treatment has immobilized the hazardous constituents of the salt solution. This paper discusses the methods used to characterize the salt solution and final Saltstone samples from 2007-2009.

  9. Ventilation System to Improve Savannah River Site's Liquid Waste...

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

    Ventilation System to Improve Savannah River Site's Liquid Waste Operations Ventilation System to Improve Savannah River Site's Liquid Waste Operations August 28, 2014 - 12:00pm...

  10. Savannah River Site's Liquid Waste Operations Adds Multi-Functional...

    Office of Environmental Management (EM)

    Savannah River Site's Liquid Waste Operations Adds Multi-Functional Laboratory Savannah River Site's Liquid Waste Operations Adds Multi-Functional Laboratory January 28, 2015 -...

  11. Membrane Treatment of Liquid Salt Bearing Radioactive Wastes

    SciTech Connect (OSTI)

    Dmitriev, S. A.; Adamovich, D. V.; Demkin, V. I.; Timofeev, E. M.

    2003-02-25T23:59:59.000Z

    The main fields of introduction and application of membrane methods for preliminary treatment and processing salt liquid radioactive waste (SLRW) can be nuclear power stations (NPP) and enterprises on atomic submarines (AS) utilization. Unlike the earlier developed technology for the liquid salt bearing radioactive waste decontamination and concentrating this report presents the new enhanced membrane technology for the liquid salt bearing radioactive waste processing based on the state-of-the-art membrane unit design, namely, the filtering units equipped with the metal-ceramic membranes of ''TruMem'' brand, as well as the electrodialysis and electroosmosis concentrators. Application of the above mentioned units in conjunction with the pulse pole changer will allow the marked increase of the radioactive waste concentrating factor and the significant reduction of the waste volume intended for conversion into monolith and disposal. Besides, the application of the electrodialysis units loaded with an ion exchange material at the end polishing stage of the radioactive waste decontamination process will allow the reagent-free radioactive waste treatment that meets the standards set for the release of the decontaminated liquid radioactive waste effluents into the natural reservoirs of fish-farming value.

  12. Decommissioning and Dismantling of Liquid Waste Storage and Liquid Waste Treatment Facility from Paldiski Nuclear Site, Estonia

    SciTech Connect (OSTI)

    Varvas, M. [AS ALARA, Leetse tee 21, Paldiski, 76806 (Estonia); Putnik, H. [Delegation of the European Commission to Russia, Kadashevskaja nab. 14/1 119017 Moscow (Russian Federation); Nirvin, B.; Pettersson, S. [SKB, Box 5864, Stockholm, SE-102 40 (Sweden); Johnsson, B. [Studsvik RadWaste, Nykoping, SE-611 82 (Sweden)

    2006-07-01T23:59:59.000Z

    The Paldiski Nuclear Facility in Estonia, with two nuclear reactors was owned by the Soviet Navy and was used for training the navy personnel to operate submarine nuclear reactors. After collapse of Soviet Union the Facility was shut down and handed over to the Estonian government in 1995. In co-operation with the Paldiski International Expert Reference Group (PIERG) decommission strategy was worked out and started to implement. Conditioning of solid and liquid operational waste and dismantling of contaminated installations and buildings were among the key issues of the Strategy. Most of the liquid waste volume, remained at the Facility, was processed in the frames of an Estonian-Finnish co-operation project using a mobile wastewater purification unit NURES (IVO International OY) and water was discharged prior to the site take-over. In 1999-2002 ca 120 m{sup 3} of semi-liquid tank sediments (a mixture of ion exchange resins, sand filters, evaporator and flocculation slurry), remained after treatment of liquid waste were solidified in steel containers and stored into interim storage. The project was carried out under the Swedish - Estonian co-operation program on radiation protection and nuclear safety. Contaminated installations in buildings, used for treatment and storage of liquid waste (Liquid Waste Treatment Facility and Liquid Waste Storage) were then dismantled and the buildings demolished in 2001-2004. (authors)

  13. ICPP radioactive liquid and calcine waste technologies evaluation. Interim report

    SciTech Connect (OSTI)

    Murphy, J.A.; Pincock, L.F.; Christiansen, I.N.

    1994-06-01T23:59:59.000Z

    The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage since 1951 and reprocessing since 1953. Until recently, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, changing world events have raised questions concerning the need to recover and recycle this material. In April 1992, DOE chose to discontinue reprocessing SNF for uranium recovery and shifted its focus toward the management and disposition of radioactive wastes accumulated through reprocessing activities. Currently, 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) and 3,800 cubic meters (m{sup 3}) of calcine waste are in inventory at the ICPP. Legal drivers and agreements exist obligating the INEL to develop, demonstrate, and implement technologies for safe and environmentally sound treatment and interim storage of radioactive liquid and calcine waste. Candidate treatment processes and waste forms are being evaluated using the Technology Evaluation and Analysis Methodology (TEAM) Model. This process allows decision makers to (1) identify optimum radioactive waste treatment and disposal form alternatives; (2) assess tradeoffs between various optimization criteria; (3) identify uncertainties in performance parameters; and (4) focus development efforts on options that best satisfy stakeholder concerns. The Systems Analysis technology evaluation presented in this document supports the DOE in selecting the most effective radioactive liquid and calcine waste management plan to implement in compliance with established regulations, court orders, and agreements.

  14. Numerical simulation of hydrothermal salt separation process and analysis and cost estimating of shipboard liquid waste disposal

    E-Print Network [OSTI]

    Hunt, Andrew Robert

    2007-01-01T23:59:59.000Z

    Due to environmental regulations, waste water disposal for US Navy ships has become a requirement which impacts both operations and the US Navy's budget. In 2006, the cost for waste water disposal Navy-wide was 54 million ...

  15. Waste Form Development for the Solidification of PDCF/MOX Liquid Waste Streams

    SciTech Connect (OSTI)

    COZZI, ALEX

    2004-02-18T23:59:59.000Z

    At the Savannah River Site, part of the Department of Energy's nuclear materials complex located in South Carolina, cementation has been selected as the solidification method for high-alpha and low-activity waste streams generated in the planned plutonium disposition facilities. A Waste Solidification Building (WSB) that will be used to treat and solidify three radioactive liquid waste streams generated by the Pit Disassembly and Conversion Facility) and the Mixed Oxide Fuel Fabrication Facility is in the preliminary design stage. The WSB is expected to treat a transuranic (TRU) waste stream composed primarily of americium and two low-level waste (LLW) streams. The acidic wastes will be concentrated in the WSB evaporator and neutralized in a cement head tank prior to solidification. A series of TRU mixes were prepared to produce waste forms exhibiting a range of processing and cured properties. The LLW mixes were prepared using the premix from the preferred TRU waste form. All of the waste forms tested passed the Toxicity Characteristic Leaching Procedure. After processing in the WSB, current plans are to dispose of the solidified TRU waste at the Waste Isolation Pilot Plant in New Mexico and the solidified LLW waste at an approved low-level waste disposal facility.

  16. Municipal Solid Waste (MSW) to Liquid Fuels Synthesis, Volume...

    Office of Environmental Management (EM)

    1: Availability of Feedstock and Technology Municipal Solid Waste (MSW) to Liquid Fuels Synthesis, Volume 1: Availability of Feedstock and Technology Municipal solid waste (MSW) is...

  17. Hydrothermal Processing of Wet Wastes

    Broader source: Energy.gov [DOE]

    Breakout Session 3A—Conversion Technologies III: Energy from Our Waste—Will we Be Rich in Fuel or Knee Deep in Trash by 2025? Hydrothermal Processing of Wet Wastes James R. Oyler, President, Genifuel Corporation

  18. 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)

    1987-01-01T23:59:59.000Z

    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.

  19. 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)

    1997-01-01T23:59:59.000Z

    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.

  20. Extraction processes and solvents for recovery of cesium, strontium, rare earth elements, technetium and actinides from liquid radioactive waste

    DOE Patents [OSTI]

    Zaitsev, Boris N. (St. Petersburg, RU); Esimantovskiy, Vyacheslav M. (St. Petersburg, RU); Lazarev, Leonard N. (St. Petersburg, RU); Dzekun, Evgeniy G. (Ozersk, RU); Romanovskiy, Valeriy N. (St. Petersburg, RU); Todd, Terry A. (Aberdeen, ID); Brewer, Ken N. (Arco, ID); Herbst, Ronald S. (Idaho Falls, ID); Law, Jack D. (Pocatello, ID)

    2001-01-01T23:59:59.000Z

    Cesium and strontium are extracted from aqueous acidic radioactive waste containing rare earth elements, technetium and actinides, by contacting the waste with a composition of a complex organoboron compound and polyethylene glycol in an organofluorine diluent mixture. In a preferred embodiment the complex organoboron compound is chlorinated cobalt dicarbollide, the polyethylene glycol has the formula RC.sub.6 H.sub.4 (OCH.sub.2 CH.sub.2).sub.n OH, and the organofluorine diluent is a mixture of bis-tetrafluoropropyl ether of diethylene glycol with at least one of bis-tetrafluoropropyl ether of ethylene glycol and bis-tetrafluoropropyl formal. The rare earths, technetium and the actinides (especially uranium, plutonium and americium), are extracted from the aqueous phase using a phosphine oxide in a hydrocarbon diluent, and reextracted from the resulting organic phase into an aqueous phase by using a suitable strip reagent.

  1. Radioactive waste processing apparatus

    DOE Patents [OSTI]

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

    1985-08-30T23:59:59.000Z

    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.

  2. Using benchmarking to minimize common DOE waste streams. Volume 1, Methodology and liquid photographic waste

    SciTech Connect (OSTI)

    Levin, V.

    1994-04-01T23:59:59.000Z

    Finding innovative ways to reduce waste streams generated at Department of Energy (DOE) sites by 50% by the year 2000 is a challenge for DOE`s waste minimization efforts. This report examines the usefulness of benchmarking as a waste minimization tool, specifically regarding common waste streams at DOE sites. A team of process experts from a variety of sites, a project leader, and benchmarking consultants completed the project with management support provided by the Waste Minimization Division EM-352. Using a 12-step benchmarking process, the team examined current waste minimization processes for liquid photographic waste used at their sites and used telephone and written questionnaires to find ``best-in-class`` industrv partners willing to share information about their best waste minimization techniques and technologies through a site visit. Eastman Kodak Co., and Johnson Space Center/National Aeronautics and Space Administration (NASA) agreed to be partners. The site visits yielded strategies for source reduction, recycle/recovery of components, regeneration/reuse of solutions, and treatment of residuals, as well as best management practices. An additional benefit of the work was the opportunity for DOE process experts to network and exchange ideas with their peers at similar sites.

  3. Heterogeneous waste processing

    DOE Patents [OSTI]

    Vanderberg, Laura A. (Los Alamos, NM); Sauer, Nancy N. (Los Alamos, NM); Brainard, James R. (Los Alamos, NM); Foreman, Trudi M. (Los Alamos, NM); Hanners, John L. (Los Alamos, NM)

    2000-01-01T23:59:59.000Z

    A combination of treatment methods are provided for treatment of heterogeneous waste including: (1) treatment for any organic compounds present; (2) removal of metals from the waste; and, (3) bulk volume reduction, with at least two of the three treatment methods employed and all three treatment methods emplyed where suitable.

  4. SUCCESSES AND EMERGING ISSUES IN SIMULATING THE PROCESSING BEHAVIOR OF LIQUID-PARTICLE NUCLEAR WASTE SLURRIES AT THE SAVANNAH RIVER SITE - 205E

    SciTech Connect (OSTI)

    Koopman, D.; Lambert, D.; Stone, M.

    2009-09-02T23:59:59.000Z

    Slurries of inorganic solids, containing both stable and radioactive elements, were produced during the cold war as by-products of the production of plutonium and enriched uranium and stored in large tanks at the Savannah River Site. Some of this high level waste is being processed into a stable glass waste form today. Waste processing involves various large scale operations such as tank mixing, inter-tank transfers, washing, gravity settling and decanting, chemical adjustment, and vitrification. The rheological properties of waste slurries are of particular interest. Methods for modeling flow curve data and predicting the properties of slurry blends are particularly important during certain operational phases. Several methods have been evaluated to predict the rheological properties of sludge slurry blends from the data on the individual slurries. These have been relatively successful.

  5. Characterization of industrial process waste heat and input heat streams

    SciTech Connect (OSTI)

    Wilfert, G.L.; Huber, H.B.; Dodge, R.E.; Garrett-Price, B.A.; Fassbender, L.L.; Griffin, E.A.; Brown, D.R.; Moore, N.L.

    1984-05-01T23:59:59.000Z

    The nature and extent of industrial waste heat associated with the manufacturing sector of the US economy are identified. Industry energy information is reviewed and the energy content in waste heat streams emanating from 108 energy-intensive industrial processes is estimated. Generic types of process equipment are identified and the energy content in gaseous, liquid, and steam waste streams emanating from this equipment is evaluated. Matchups between the energy content of waste heat streams and candidate uses are identified. The resultant matrix identifies 256 source/sink (waste heat/candidate input heat) temperature combinations. (MHR)

  6. 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

    2010-01-30T23:59:59.000Z

    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.

  7. Method for processing aqueous wastes

    DOE Patents [OSTI]

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

    1993-12-28T23:59:59.000Z

    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.

  8. 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)

    1993-01-01T23:59:59.000Z

    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.

  9. Independent Oversight Assessment, Salt Waste Processing Facility...

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

    Salt Waste Processing Facility Project - January 2013 January 2013 Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project The U.S. Department...

  10. A Regulatory Analysis and Reassessment of U.S. Environmental Protection Agency Listed Hazardous Waste Numbers for Applicability to the INTEC Liquid Waste System

    SciTech Connect (OSTI)

    Gilbert, K.L.; Venneman, T.E.

    1998-12-01T23:59:59.000Z

    This report concludes that there are four listed hazardous waste numbers (F001, F002, F005, and U134) applicable to the waste in the Process Equipment Waste Evaporator (PEWE) liquid waste system at the Idaho National Engineering and Environmental Laboratory. The chemical constituents associated with these listed hazardous waste numbers, including those listed only for ignitability are identified. The RCRA Part A permit application hazardous waste numbers identify chemical constituents that may be treated or stored by the PEWE liquid waste system either as a result of a particular characteristic (40 CFR, Subpart C) or as a result of a specific process (40 CFR 261, Subpart D). The RCRA Part A permit application for the PEWE liquid waste system identifies the universe of Environmental Protection Agency (EPA) hazardous waste numbers [23 characteristic (hazardous waste codes) numbers and 105 listed numbers (four F-listed hazardous waste numbers, 20 P-listed hazardous waste numbers, and 81 U-listed hazardous waste numbers)] deemed acceptable for storage and treatment. This evaluation, however, identifies only listed wastes (and their chemical constituents) that have actually entered the PEWE liquid waste system and would, therefore, be assigned to the PEWE liquids and treatment residuals.

  11. Future radioactive liquid waste streams study

    SciTech Connect (OSTI)

    Rey, A.S.

    1993-11-01T23:59:59.000Z

    This study provides design planning information for the Radioactive Liquid Waste Treatment Facility (RLWTF). Predictions of estimated quantities of Radioactive Liquid Waste (RLW) and radioactivity levels of RLW to be generated are provided. This information will help assure that the new treatment facility is designed with the capacity to treat generated RLW during the years of operation. The proposed startup date for the RLWTF is estimated to be between 2002 and 2005, and the life span of the facility is estimated to be 40 years. The policies and requirements driving the replacement of the current RLW treatment facility are reviewed. Historical and current status of RLW generation at Los Alamos National Laboratory are provided. Laboratory Managers were interviewed to obtain their insights into future RLW activities at Los Alamos that might affect the amount of RLW generated at the Lab. Interviews, trends, and investigation data are analyzed and used to create scenarios. These scenarios form the basis for the predictions of future RLW generation and the level of RLW treatment capacity which will be needed at LANL.

  12. Potential radiation damage: Storage tanks for liquid radioactive waste

    SciTech Connect (OSTI)

    Caskey, G.R. Jr.

    1992-08-21T23:59:59.000Z

    High level waste at SRS is stored in carbon steel tanks constructed during the period 1951 to 1981. This waste contains radionuclides that decay by alpha, beta, or gamma emission or are spontaneous neutronsources. Thus, a low intensity radiation field is generated that is capable of causing displacement damage to the carbon steel. The potential for degradation of mechanical properties was evaluated by comparing the estimated displacement damage with published data relating changes in Charpy V-notch (CVN) impact energy to neutron exposure. Experimental radiation data was available for three of the four grades of carbonsteel from which the tanks were constructed and is applicable to all four steels. Estimates of displacement damage arising from gamma and neutron radiation have been made based on the radionuclide contents for high level waste that are cited in the Safety Analysis Report (SAR) for the Liquid Waste Handling Facilities in the 200-Area. Alpha and beta emissions do not penetrate carbon steel to a sufficient depth to affect the bulk properties of the tank walls but may aggravate corrosion processes. The damage estimates take into account the source of the waste (F- or H-Area), the several types of tank service, and assume wateras an attenuating medium. Estimates of displacement damage are conservative because they are based on the highest levels of radionuclide contents reported in the SAR and continuous replenishment of the radionuclides.

  13. Consolidation process for producing ceramic waste forms

    DOE Patents [OSTI]

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

    2000-01-01T23:59:59.000Z

    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.

  14. Detection of free liquid in containers of solidified radioactive waste

    DOE Patents [OSTI]

    Greenhalgh, Wilbur O. (Richland, WA)

    1985-01-01T23:59:59.000Z

    A method of nondestructively detecting the presence of free liquid within a sealed enclosure containing solidified waste by measuring the levels of waste at two diametrically opposite locations while slowly tilting the enclosure toward one of said locations. When the measured level remains constant at the other location, the measured level at said one location is noted and any measured difference of levels indicates the presence of liquid on the surface of the solidified waste. The absence of liquid in the enclosure is verified when the measured levels at both locations are equal.

  15. NOCHAR Polymers: An Aqueous and Organic Liquid Solidification Process for Cadarache LOR (Liquides Organiques Radioactifs) - 13195

    SciTech Connect (OSTI)

    Vaudey, Claire-Emilie; Renou, Sebastien; Porco, Julien [AREVA CL BU, STMI ZAC de Courcelle 1 route de la Noue 91196 Gif-sur-Yvette (France)] [AREVA CL BU, STMI ZAC de Courcelle 1 route de la Noue 91196 Gif-sur-Yvette (France); Kelley, Dennis [Pacific World Trade, Inc. Hillsdale Technical Center 6970 Hillsdale Court Indianapolis, Indiana 46250 (United States)] [Pacific World Trade, Inc. Hillsdale Technical Center 6970 Hillsdale Court Indianapolis, Indiana 46250 (United States); Cochaud, Chantal [CEA Cadarache, DSN/SGTD, 13108 St Paul Lez Durance Cedex (France)] [CEA Cadarache, DSN/SGTD, 13108 St Paul Lez Durance Cedex (France); Serrano, Roger [CEA Marcoule, DPAD, BP 17171, 30207 Bagnols sur Ceze Cedex (France)] [CEA Marcoule, DPAD, BP 17171, 30207 Bagnols sur Ceze Cedex (France)

    2013-07-01T23:59:59.000Z

    To handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW) in France, two options can be considered: the incineration at CENTRACO facility and the disposal facility on ANDRA sites. The waste acceptance in these radwaste routes is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the radwaste route specifications. If the waste characteristics are incompatible with the radwaste route specifications (presence of significant quantities of chlorine, fluorine, organic component etc or/and high activity limits), it is necessary to find an alternative solution that consists of a waste pre-treatment process. In the context of the problematic Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. The first one is composed of organic liquids at 13.1 % (diphenyloxazol, mesitylene, TBP, xylene) and water at 86.9 %. The second one is composed of TBP at 8.6 % and water at 91.4 %. They contain chlorine, fluorine and sulphate and have got alpha/beta/gamma spectra with mass activities equal to some kBq.g{sup -1}. Therefore, tritium is present and creates the second problematic waste stream. As a consequence, in order for disposal acceptance at the ANDRA site, it is necessary to pre-treat the waste. The NOCHAR polymers as an aqueous and organic liquid solidification process seem to be an adequate solution. Indeed, these polymers constitute an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing etc) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and immobilise the liquid. Then as the polymer-organic cure, over time, the polymer continues to collapse on the organic to create a permanent bond. The N960 has the ability to absorb aqueous waste up to 100 times its own weight. It creates a strong mechanical bond which permanently traps the contamination imbedded in the aqueous liquids. As a consequence, these two polymers seem to be able to constitute a suitable solidification matrix for a final acceptance in storage on ANDRA sites. In order to validate the polymers as an acceptable aqueous and organic solidification process for Cadarache LOR, some solidification tests realized with N910 and N960, have been carried out for different Waste/Polymer ratios. The determination of the best Waste/Polymer ratio and of the optimal experimental parameters has been made through an exudation test. Indeed, the process prevents leaching and it results in the absence of residual free organic or aqueous liquid which is forbidden in storage by ANDRA specifications. With these test results, we generated scientific data which are fundamental to obtain an ANDRA agreement. As a conclusion, the aim of this study is to demonstrate that the solidification by polymers can constitute a pre-treatment solution for Cadarache LOR and more generally, for various organic and mixed organic/aqueous waste which can not be directly acceptable at CENTRACO facility or at ANDRA storage sites. This work is, therefore, a solid background to demonstrate the feasibility of the waste pre-treatment by solidification with polymers and to encourage the development of this process. (authors)

  16. Municipal Solid Waste (MSW) to Liquid Fuels Synthesis, Volume...

    Energy Savers [EERE]

    2: A Techno-economic Evaluation of the Production of Mixed Alcohols Municipal Solid Waste (MSW) to Liquid Fuels Synthesis, Volume 2: A Techno-economic Evaluation of the Production...

  17. Technical evaluation of proposed Ukrainian Central Radioactive Waste Processing Facility

    SciTech Connect (OSTI)

    Gates, R.; Glukhov, A.; Markowski, F.

    1996-06-01T23:59:59.000Z

    This technical report is a comprehensive evaluation of the proposal by the Ukrainian State Committee on Nuclear Power Utilization to create a central facility for radioactive waste (not spent fuel) processing. The central facility is intended to process liquid and solid radioactive wastes generated from all of the Ukrainian nuclear power plants and the waste generated as a result of Chernobyl 1, 2 and 3 decommissioning efforts. In addition, this report provides general information on the quantity and total activity of radioactive waste in the 30-km Zone and the Sarcophagus from the Chernobyl accident. Processing options are described that may ultimately be used in the long-term disposal of selected 30-km Zone and Sarcophagus wastes. A detailed report on the issues concerning the construction of a Ukrainian Central Radioactive Waste Processing Facility (CRWPF) from the Ukrainian Scientific Research and Design institute for Industrial Technology was obtained and incorporated into this report. This report outlines various processing options, their associated costs and construction schedules, which can be applied to solving the operating and decommissioning radioactive waste management problems in Ukraine. The costs and schedules are best estimates based upon the most current US industry practice and vendor information. This report focuses primarily on the handling and processing of what is defined in the US as low-level radioactive wastes.

  18. Municipal Solid Waste (MSW) to Liquid Fuels Synthesis, Volume 1: Availability of Feedstock and Technology

    SciTech Connect (OSTI)

    Valkenburt, Corinne; Walton, Christie W.; Thompson, Becky L.; Gerber, Mark A.; Jones, Susanne B.; Stevens, Don J.

    2008-12-01T23:59:59.000Z

    This report investigated the potential of using municipal solid waste (MSW) to make synthesis gas (syngas) suitable for production of liquid fuels. Issues examined include: • MSW physical and chemical properties affecting its suitability as a gasifier feedstock and for liquid fuels synthesis • expected process scale required for favorable economics • the availability of MSW in quantities sufficient to meet process scale requirements • the state-of-the-art of MSW gasification technology.

  19. Liquid and Gaseous Waste Operations Project Annual Operating Report CY 1999

    SciTech Connect (OSTI)

    Maddox, J.J.; Scott, C.B.

    2000-03-01T23:59:59.000Z

    A total of 5.77 x 10 7 gallons (gal) of liquid waste was decontaminated by the Process Waste Treatment Complex (PWTC) - Building 3544 ion exchange system during calendar year (CY) 1999. This averaged to 110 gpm throughout the year. An additional 3.94 x 10 6 gal of liquid waste (average of 8 gpm throughout the year) was decontaminated using the zeolite treatment system due to periods of high Cesium levels in the influent wastewater. A total of 6.17 x 10 7 gal of liquid waste (average of 118 gpm throughout the year) was decontaminated at Building 3544 during the year. During the year, the regeneration of the ion exchange resins resulted in the generation of 8.00 x 10 3 gal of Liquid Low-Level Waste (LLLW) concentrate and 9.00 x 10 2 gal of LLLW supernate. See Table 1 for a monthly summary of activities at Building 3544. Figure 1 shows a diagram of the Process Waste Collection and Transfer System and Figure 2 shows a diagram of the Building 3544 treatment process. Figures 3, 4 5, and 6 s how a comparison of operations at Building 3544 in 1997 with previous years. Figure 7 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1995.

  20. Solvent extraction in the treatment of acidic high-level liquid waste : where do we stand?

    SciTech Connect (OSTI)

    Horwitz, E. P.; Schulz, W. W.

    1998-06-18T23:59:59.000Z

    During the last 15 years, a number of solvent extraction/recovery processes have been developed for the removal of the transuranic elements, {sup 90}Sr and {sup 137}Cs from acidic high-level liquid waste. These processes are based on the use of a variety of both acidic and neutral extractants. This chapter will present an overview and analysis of the various extractants and flowsheets developed to treat acidic high-level liquid waste streams. The advantages and disadvantages of each extractant along with comparisons of the individual systems are discussed.

  1. Treatment of Bottled Liquid Waste During Remediation of the Hanford 618-10 Burial Ground - 13001

    SciTech Connect (OSTI)

    Faulk, Darrin E.; Pearson, Chris M.; Vedder, Barry L.; Martin, David W. [Washington Closure Hanford, LLC, Richland, WA 99354 (United States)] [Washington Closure Hanford, LLC, Richland, WA 99354 (United States)

    2013-07-01T23:59:59.000Z

    A problematic waste form encountered during remediation of the Hanford Site 618-10 burial ground consists of bottled aqueous waste potentially contaminated with regulated metals. The liquid waste requires stabilization prior to landfill disposal. Prior remediation activities at other Hanford burial grounds resulted in a standard process for sampling and analyzing liquid waste using manual methods. Due to the highly dispersible characteristics of alpha contamination, and the potential for shock sensitive chemicals, a different method for bottle processing was needed for the 618-10 burial ground. Discussions with the United States Department of Energy (DOE) and United States Environmental Protection Agency (EPA) led to development of a modified approach. The modified approach involves treatment of liquid waste in bottles, up to one gallon per bottle, in a tray or box within the excavation of the remediation site. Bottles are placed in the box, covered with soil and fixative, crushed, and mixed with a Portland cement grout. The potential hazards of the liquid waste preclude sampling prior to treatment. Post treatment verification sampling is performed to demonstrate compliance with land disposal restrictions and disposal facility acceptance criteria. (authors)

  2. Process for remediation of plastic waste

    DOE Patents [OSTI]

    Pol, Vilas G; Thiyagarajan, Pappannan

    2013-11-12T23:59:59.000Z

    A single step process for degrading plastic waste by converting the plastic waste into carbonaceous products via thermal decomposition of the plastic waste by placing the plastic waste into a reactor, heating the plastic waste under an inert or air atmosphere until the temperature of about 700.degree. C. is achieved, allowing the reactor to cool down, and recovering the resulting decomposition products therefrom. The decomposition products that this process yields are carbonaceous materials, and more specifically carbon nanotubes having a partially filled core (encapsulated) adjacent to one end of the nanotube. Additionally, in the presence of a transition metal compound, this thermal decomposition process produces multi-walled carbon nanotubes.

  3. Process for remediation of plastic waste

    DOE Patents [OSTI]

    Pol, Vilas G. (Westmont, IL); Thiyagarajan, Pappannan (Germantown, MD)

    2012-04-10T23:59:59.000Z

    A single step process for degrading plastic waste by converting the plastic waste into carbonaceous products via thermal decomposition of the plastic waste by placing the plastic waste into a reactor, heating the plastic waste under an inert or air atmosphere until the temperature of 700.degree. C. is achieved, allowing the reactor to cool down, and recovering the resulting decomposition products therefrom. The decomposition products that this process yields are carbonaceous materials, and more specifically egg-shaped and spherical-shaped solid carbons. Additionally, in the presence of a transition metal compound, this thermal decomposition process produces multi-walled carbon nanotubes.

  4. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report

    SciTech Connect (OSTI)

    Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Argyle, Mark Don; Lauerhass, Lance; Bendixsen, Carl Lee; Hinckley, Steve Harold

    2000-11-01T23:59:59.000Z

    The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

  5. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report

    SciTech Connect (OSTI)

    Herbst, A.K.; McCray, J.A.; Kirkham, R.J.; Pao, J.; Argyle, M.D.; Lauerhass, L.; Bendixsen, C.L.; Hinckley, S.H.

    2000-10-31T23:59:59.000Z

    The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

  6. Transuranic (TRU) Waste Processing Center- Overview

    Broader source: Energy.gov [DOE]

    DOE established the TRU Waste Processing Center (TWPC) as a regional center for the management, treatment, packaging and shipment of DOE TRU waste legacy inventory. TWPC is also responsible for managing and treating Low Level and Mixed Low Level Waste generated at ORNL. TWPC is operated by Wastren Advantage, Inc. (WAI) under contract to the DOE's Oak Ridge Office.

  7. CMD-3 Liquid Xenon Calorimeter's signals processing

    E-Print Network [OSTI]

    CMD-3 Liquid Xenon Calorimeter's signals processing for timing measurements. Leonid Epshtein Budker connected to constitute 264 «towers»; signal of each tower is processed by electronic channel. Liquid Xenon functional channel scheme and signal's shapes in the typical point. Digital Signal Processing ADCCSALXe

  8. Statistical Description of Liquid Low-Level Waste System Transssuranic Wastes at Oak Ridge Nation Laboratory, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    NONE

    1996-12-01T23:59:59.000Z

    The US DOE has presented plans for processing liquid low-level wastes (LLLW) located at Oak Ridge National Laboratory (ORNL) in the LLLW tank system. These wastes are among the most hazardous on the Oak Ridge reservation and exhibit both RCRA toxic and radiological hazards. The Tennessee Department of Health and Environment has mandated that the processing of these wastes must begin by the year 2002 and the the goal should be permanent disposal at a site off the Oak Ridge Reservation. To meet this schedule, DOE will solicit bids from various private sector companies for the construction of a processing facility on land located near the ORNL Melton Valley Storage Tanks to be operated by the private sector on a contract basis. This report will support the Request for Proposal process and will give potential vendors information about the wastes contained in the ORNL tank farm system. The report consolidates current data about the properties and composition of these wastes and presents methods to calculate the error bounds of the data in the best technically defensible manner possible. The report includes information for only the tank waste that is to be included in the request for proposal.

  9. Liquid waste certification plan 340 waste handling facility

    SciTech Connect (OSTI)

    HALGREN, D.L.

    1999-04-21T23:59:59.000Z

    This document addresses the discharges from the 340 Facility to the 300 Area Process Sewer and Retention Process Sewer.

  10. Iraq liquid radioactive waste tanks maintenance and monitoring program plan.

    SciTech Connect (OSTI)

    Dennis, Matthew L.; Cochran, John Russell; Sol Shamsaldin, Emad (Iraq Ministry of Science and Technology)

    2011-10-01T23:59:59.000Z

    The purpose of this report is to develop a project management plan for maintaining and monitoring liquid radioactive waste tanks at Iraq's Al-Tuwaitha Nuclear Research Center. Based on information from several sources, the Al-Tuwaitha site has approximately 30 waste tanks that contain varying amounts of liquid or sludge radioactive waste. All of the tanks have been non-operational for over 20 years and most have limited characterization. The program plan embodied in this document provides guidance on conducting radiological surveys, posting radiation control areas and controlling access, performing tank hazard assessments to remove debris and gain access, and conducting routine tank inspections. This program plan provides general advice on how to sample and characterize tank contents, and how to prioritize tanks for soil sampling and borehole monitoring.

  11. Process for recovery of palladium from nuclear fuel reprocessing wastes

    DOE Patents [OSTI]

    Campbell, D.O.; Buxton, S.R.

    1980-06-16T23:59:59.000Z

    Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M; (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound; (c) heating the solution at reflux temperature until precipitation is complete; and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

  12. Method for co-processing waste rubber and carbonaceous material

    DOE Patents [OSTI]

    Farcasiu, Malvina (Pittsburgh, PA); Smith, Charlene M. (Pittsburgh, PA)

    1991-01-01T23:59:59.000Z

    In a process for the co-processing of waste rubber and carbonaceous material to form a useful liquid product, the rubber and the carbonaceous material are combined and heated to the depolymerization temperature of the rubber in the presence of a source of hydrogen. The depolymerized rubber acts as a liquefying solvent for the carbonaceous material while a beneficial catalytic effect is obtained from the carbon black released on depolymerization the reinforced rubber. The reaction is carried out at liquefaction conditions of 380.degree.-600.degree. C. and 70-280 atmospheres hydrogen pressure. The resulting liquid is separated from residual solids and further processed such as by distillation or solvent extraction to provide a carbonaceous liquid useful for fuels and other purposes.

  13. Method for co-processing waste rubber and carbonaceous material

    SciTech Connect (OSTI)

    Farcasiu, M.; Smith, C.M.

    1990-10-09T23:59:59.000Z

    In a process for the co-processing of waste rubber and carbonaceous material to form a useful liquid product, the rubber and the carbonaceous material are combined and heated to the depolymerization temperature of the rubber in the presence of a source of hydrogen. The deploymerized rubber acts as a liquefying solvent for the carbonaceous material while a beneficial catalytic effect is obtained from the carbon black released on deploymerization the reinforced rubber. The reaction is carried out at liquefaction conditions of 380--600{degrees}C and 70--280 atmospheres hydrogen pressure. The resulting liquid is separated from residual solids and further processed such as by distillation or solvent extraction to provide a carbonaceous liquid useful for fuels and other purposes.

  14. Decontamination of Nuclear Liquid Wastes Status of CEA and AREVA R and D: Application to Fukushima Waste Waters - 12312

    SciTech Connect (OSTI)

    Fournel, B.; Barre, Y.; Lepeytre, C.; Peycelon, H. [CEA Marcoule, DTCD, BP17171, 30207 Bagnols sur Ceze (France); Grandjean, A. [Institut de Chimie Separative de Marcoule, UMR5257 CEA-CNRS-UM2-ENSCM, BP17171, 30207 Bagnols sur Ceze (France); Prevost, T.; Valery, J.F. [AREVA NC, Paris La Defense (France); Shilova, E.; Viel, P. [CEA Saclay, DSM/IRAMIS/SPCSI, 91191 Gif sur Yvette (France)

    2012-07-01T23:59:59.000Z

    Liquid wastes decontamination processes are mainly based on two techniques: Bulk processes and the so called Cartridges processes. The first technique has been developed for the French nuclear fuel reprocessing industry since the 60's in Marcoule and La Hague. It is a proven and mature technology which has been successfully and quickly implemented by AREVA at Fukushima site for the processing of contaminated waters. The second technique, involving cartridges processes, offers new opportunities for the use of innovative adsorbents. The AREVA process developed for Fukushima and some results obtained on site will be presented as well as laboratory scale results obtained in CEA laboratories. Examples of new adsorbents development for liquid wastes decontamination are also given. A chemical process unit based on co-precipitation technique has been successfully and quickly implemented by AREVA at Fukushima site for the processing of contaminated waters. The asset of this technique is its ability to process large volumes in a continuous mode. Several chemical products can be used to address specific radioelements such as: Cs, Sr, Ru. Its drawback is the production of sludge (about 1% in volume of initial liquid volume). CEA developed strategies to model the co-precipitation phenomena in order to firstly minimize the quantity of added chemical reactants and secondly, minimize the size of co-precipitation units. We are on the way to design compact units that could be mobilized very quickly and efficiently in case of an accidental situation. Addressing the problem of sludge conditioning, cementation appears to be a very attractive solution. Fukushima accident has focused attention on optimizations that should be taken into account in future studies: - To better take account for non-typical aqueous matrixes like seawater; - To enlarge the spectrum of radioelements that can be efficiently processed and especially short lives radioelements that are usually less present in standard effluents resulting from nuclear activities; - To develop reversible solid adsorbents for cartridge-type applications in order to minimize wastes. (authors)

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

    SciTech Connect (OSTI)

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

    2014-12-01T23:59:59.000Z

    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.

  16. Voluntary Protection Program Onsite Review, Salt Waste Processing...

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

    Salt Waste Processing Facility Construction Project - February 2013 Voluntary Protection Program Onsite Review, Salt Waste Processing Facility Construction Project - February 2013...

  17. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report

    SciTech Connect (OSTI)

    Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Hinckley, Steve Harold

    1999-10-01T23:59:59.000Z

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

  18. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report

    SciTech Connect (OSTI)

    A. K. Herbst; J. A. McCray; R. J. Kirkham; J. Pao; S. H. Hinckley

    1999-09-30T23:59:59.000Z

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

  19. Tank waste remediation system high-level waste feed processability assessment report

    SciTech Connect (OSTI)

    Lambert, S.L. [Westinghouse Hanford Co., Richland, WA (United States); Kim, D.S. [Pacific Northwest Lab., Richland, WA (United States)

    1994-12-01T23:59:59.000Z

    This study evaluates the effect of feed composition on the performance of the high-level vitrification process. It is assumed in this study that the tank wastes are retrieved and blended by tank farms, producing 12 different blends from the single-shell tank farms, two blends of double-shell tank waste, and a separately defined all-tank blend. This blending scenario was chosen only for evaluating the impact of composition on the volume of high- level waste glass produced. Special glass compositions were formulated for each waste blend based on glass property models and the properties of similar glasses. These glasses were formulated to meet the applicable viscosity, electrical conductivity, and liquidus temperature constraints for the identified candidate melters. Candidate melters in this study include the low-temperature stirred melter, which operates at 1050{degrees}C; the reference Hanford Waste Vitrification Plant liquid-fed ceramic melter, which operates at 1150{degrees}C; and the high-temperature, joule-heated melter and the cold-crucible melter, which operate over a temperature range of 1150{degrees}C to 1400{degrees}C. In the most conservative case, it is estimated that 61,000 MT of glass will be produced if the Site`s high-level wastes are retrieved by tank farms and processed in the reference joule-heated melter. If an all-tank blend was processed under the same conditions, the reference melter would produce 21,250 MT of glass. If cross-tank blending were used, it is anticipated that $2.0 billion could be saved in repository disposal costs (based on an average disposal cost of $217,000 per canister) by blending the S, SX, B, and T Tank Farm wastes with other wastes prior to vitrification. General blending among all the tank farms is expected to produce great potential benefit.

  20. On-Site Decontamination System for Liquid Low Level Radioactive Waste - 13010

    SciTech Connect (OSTI)

    OSMANLIOGLU, Ahmet Erdal [Cekmece Nuclear Research and Training Center, Kucukcekmece Istanbul (Turkey)] [Cekmece Nuclear Research and Training Center, Kucukcekmece Istanbul (Turkey)

    2013-07-01T23:59:59.000Z

    This study is based on an evaluation of purification methods for liquid low-level radioactive waste (LLLW) by using natural zeolite. Generally the volume of liquid low-level waste is relatively large and the specific activity is rather low when compared to other radioactive waste types. In this study, a pilot scale column was used with natural zeolite as an ion exchanger media. Decontamination and minimization of LLLW especially at the generation site decrease operational cost in waste management operations. Portable pilot scale column was constructed for decontamination of LLW on site. Effect of temperature on the radionuclide adsorption of the zeolite was determined to optimize the waste solution temperature for the plant scale operations. In addition, effect of pH on the radionuclide uptake of the zeolite column was determined to optimize the waste solution pH for the plant scale operations. The advantages of this method used for the processing of LLLW are discussed in this paper. (authors)

  1. Evaluation of prospective hazardous waste treatment technologies for use in processing low-level mixed wastes at Rocky Flats

    SciTech Connect (OSTI)

    McGlochlin, S.C.; Harder, R.V.; Jensen, R.T.; Pettis, S.A.; Roggenthen, D.K.

    1990-09-18T23:59:59.000Z

    Several technologies for destroying or decontaminating hazardous wastes were evaluated (during early 1988) as potential processes for treating low-level mixed wastes destined for destruction in the Fluidized Bed Incinerator. The processes that showed promise were retained for further consideration and placed into one (or more) of three categories based on projected availability: short, intermediate, and long-term. Three potential short-term options were identified for managing low-level mixed wastes generated or stored at the Rocky Flats Plant (operated by Rockwell International in 1988). These options are: (1) Continue storing at Rocky Flats, (2) Ship to Nevada Test Site for landfill disposal, or (3) Ship to the Idaho National Engineering Laboratory for incineration in the Waste Experimental Reduction Facility. The third option is preferable because the wastes will be destroyed. Idaho National Engineering Laboratory has received interim status for processing solid and liquid low-level mixed wastes. However, low-level mixed wastes will continue to be stored at Rocky Flats until the Department of Energy approval is received to ship to the Nevada Test Site or Idaho National Engineering Laboratory. Potential intermediate and long-term processes were identified; however, these processes should be combined into complete waste treatment systems'' that may serve as alternatives to the Fluidized Bed Incinerator. Waste treatment systems will be the subject of later work. 59 refs., 2 figs.

  2. Process waste assessment for solid low-level radioactive waste and solid TRU waste

    SciTech Connect (OSTI)

    Haney, L. [Westinghouse Savannah River Co., Aiken, SC (United States); Gamble, G.S. [Law Environmental, Inc., Kennesaw, GA (United States)

    1994-04-01T23:59:59.000Z

    Process Waste Assessments (PWAs) are a necessary and important part of a comprehensive waste management plan. PWAs are required by Federal RCRA regulations, certain state regulations and Department of Energy Orders. This paper describes the assessment process and provides examples used by Law Environmental, Inc., in performing numerous PWAs at the Savannah River Site in Aiken, SC.

  3. Process for treating alkaline wastes for vitrification

    DOE Patents [OSTI]

    Hsu, Chia-lin W.

    1994-01-01T23:59:59.000Z

    According to its major aspects and broadly stated, the present invention is a process for treating alkaline waste materials, including high level radioactive wastes, for vitrification. The process involves adjusting the pH of the wastes with nitric acid, adding formic acid (or a process stream containing formic acid) to reduce mercury compounds to elemental mercury and MnO{sub 2} to the Mn(II) ion, and mixing with class formers to produce a melter feed. The process minimizes production of hydrogen due to noble metal-catalyzed formic acid decomposition during, treatment, while producing a redox-balanced feed for effective melter operation and a quality glass product. An important feature of the present invention is the use of different acidifying and reducing, agents to treat the wastes. The nitric acid acidifies the wastes to improve yield stress and supplies acid for various reactions; then the formic acid reduces mercury compounds to elemental mercury and MnO{sub 2}) to the Mn(II) ion. When the pH of the waste is lower, reduction of mercury compounds and MnO{sub 2}) is faster and less formic acid is needed, and the production of hydrogen caused by catalytically-active noble metals is decreased.

  4. Exploratory study of complexant concentrate waste processing

    SciTech Connect (OSTI)

    Lumetta, G.J.; Bray, L.A.; Kurath, D.E.; Morrey, J.R.; Swanson, J.L.; Wester, D.W.

    1993-02-01T23:59:59.000Z

    The purpose of this exploratory study, conducted by Pacific Northwest Laboratory for Westinghouse Hanford Company, was to determine the effect of applying advanced chemical separations technologies to the processing and disposal of high-level wastes (HLW) stored in underground tanks. The major goals of this study were to determine (1) if the wastes can be partitioned into a small volume of HLW plus a large volume of low-level waste (LLW), and (2) if the activity in the LLW can be lowered enough to meet NRC Class LLW criteria. This report presents the results obtained in a brief scouting study of various processes for separating radionuclides from Hanford complexant concentrate (CC) waste.

  5. Examples of Process Modifications that Reduce Waste

    E-Print Network [OSTI]

    Nelson, K. E.

    : Vented to the air Discharged to the water Sent to landfill Sent to an incinerator Sent to a tlare Sent to a biological treatment facility The two most common waste reduction techniques are to make less waste initially and to recycle waste products... Industrial Energy Technology Conference, Houston, TX, June 19-20, 1990 2. Use off-spec material Occasionally. a process can use off-spec material (that would otherwise be burned or landfilled) because the particular quality that makes the material off...

  6. Industrial Technology of Decontamination of Liquid Radioactive Waste in SUE MosSIA 'Radon' - 12371

    SciTech Connect (OSTI)

    Adamovich, Dmitry V.; Neveykin, Petr P.; Karlin, Yuri V.; Savkin, Alexander E. [SUE MosSIA 'Radon', 7th Rostovsky lane 2/14, Moscow 119121 (Russian Federation)

    2012-07-01T23:59:59.000Z

    SUE MosSIA 'RADON' - this enterprise was created more than 50 years ago, which deals with the recycling of radioactive waste and conditioning of spent sources of radiation in stationary and mobile systems in the own factory and operating organizations. Here is represented the experience SUE MosSIA 'Radon' in the field of the management with liquid radioactive waste. It's shown, that the activity of SUE MosSIA 'RADON' is developing in three directions - improvement of technical facilities for treatment of radioactive waters into SUE MosSIA 'RADON' development of mobile equipment for the decontamination of radioactive waters in other organizations, development of new technologies for decontamination of liquid radioactive wastes as part of various domestic Russian and international projects including those related to the operation of nuclear power and nuclear submarines. SUE MosSIA 'RADON' has processed more than 270 thousand m{sup 3} of radioactive water, at that more than 7000 m{sup 3} in other organizations for more than 50 years. It is shown that a number of directions, particularly, the development of mobile modular units for decontamination of liquid radioactive waste, SUE MosSIA 'RADON' is a leader in the world. (authors)

  7. 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.

    1980-04-01T23:59:59.000Z

    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.

  8. Recovery and utilization of waste liquids in ultra-clean coal preparation by chemical leaching

    SciTech Connect (OSTI)

    Xu Zesheng; Shi Zhimin; Yang Qiaowen; Wang Xinguo [China Univ. of Mining and Technology, Beijing (China). Beijing Graduate School

    1997-12-31T23:59:59.000Z

    Coal with ash lower than 1%, being called an ultra-clean coal, has many potential applications, such as a substitute for diesel fuel, production of carbon electrodes, superior activated carbon and other chemical materials. It is difficult to reduce coal ash to such a level by conventional coal preparation technology. By means of chemical leaching with the proper concentration of alkali and acid solutions, any coal can be deeply deashed to 1% ash level. However, the cost of chemical methods is higher than that of physical ones, additionally, the waste liquids would give rise to environmental pollution if used on a large scale. If the waste liquids from chemical preparation of ultra-clean coal can be recovered and utilized, so as to produce salable by-products, the cost of chemical leaching will be reduced. This processing will also solve the pollution problem of these waste liquids. This paper describes recovery and utilization methods for these liquids used in chemical leaching, including the recoveries of alkali, silica, sodium-salt and aluminium-salt. A preliminary estimate was made regarding its economic benefits. It shows that this research solves the two problems in the chemical preparation of ultra-clean coal. One is the high-cost and the other is environmental pollution. This research demonstrates good potential for the production of ultra-clean coal on an industrial scale.

  9. Emissions model of waste treatment operations at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Schindler, R.E.

    1995-03-01T23:59:59.000Z

    An integrated model of the waste treatment systems at the Idaho Chemical Processing Plant (ICPP) was developed using a commercially-available process simulation software (ASPEN Plus) to calculate atmospheric emissions of hazardous chemicals for use in an application for an environmental permit to operate (PTO). The processes covered by the model are the Process Equipment Waste evaporator, High Level Liquid Waste evaporator, New Waste Calcining Facility and Liquid Effluent Treatment and Disposal facility. The processes are described along with the model and its assumptions. The model calculates emissions of NO{sub x}, CO, volatile acids, hazardous metals, and organic chemicals. Some calculated relative emissions are summarized and insights on building simulations are discussed.

  10. Solvent extraction and recovery of the transuranic elements from waste solutions using the TRUEX process

    SciTech Connect (OSTI)

    Horwitz, E.P.; Schulz, W.W.

    1985-01-01T23:59:59.000Z

    High-level liquid waste is produced during the processing of irradiated nuclear fuel by the PUREX process. In some cases the treatment of metallurgical scrap to recover the plutonium values also generates a nitric acid waste solution. Both waste solutions contain sufficient concentrations of transuranic elements (mostly /sup 241/Am) to require handling and disposal as a TRU waste. This paper describes a recently developed solvent extraction/recovery process called TRUEX (transuranium extraction) which is designed to reduce the TRU concentration in nitric waste solutions to <100 nCi/g of disposed form (1,2). (In the USA, non-TRU waste is defined as <100 nCi of TRU/g of disposed form.) The process utilizes PUREX process solvent (TBP in a normal paraffinic hydrocarbon or carbon tetrachloride) modified by a small concentration of octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (abbrev. CMPO). The presence of CMPO enables the modified PUREX process solvent to extract trivalent actinides as well as tetra- and hexavalent actinides. A major feature of the TRUEX process is that is is applicable to waste solutions containing a wide range of nitric acid, salt, and fission product concentrations and at the same time is very compatible with existing liquid-liquid extraction technology as usually practiced in a fuel reprocessing plant. To date the process has been tested on two different types of synthetic waste solutions. The first solution is a typical high-level nitric acid waste and the second a typical waste solution generated in metallurgical scrap processing. Results are discussed. 4 refs., 1 fig., 4 tabs.

  11. Process for the production of liquid hydrocarbons

    DOE Patents [OSTI]

    Bhatt, Bharat Lajjaram; Engel, Dirk Coenraad; Heydorn, Edward Clyde; Senden, Matthijis Maria Gerardus

    2006-06-27T23:59:59.000Z

    The present invention concerns a process for the preparation of liquid hydrocarbons which process comprises contacting synthesis gas with a slurry of solid catalyst particles and a liquid in a reactor vessel by introducing the synthesis gas at a low level into the slurry at conditions suitable for conversion of the synthesis gas into liquid hydrocarbons, the solid catalyst particles comprising a catalytic active metal selected from cobalt or iron on a porous refractory oxide carrier, preferably selected from silica, alumina, titania, zirconia or mixtures thereof, the catalyst being present in an amount between 10 and 40 vol. percent based on total slurry volume liquids and solids, and separating liquid material from the solid catalyst particles by using a filtration system comprising an asymmetric filtration medium (the selective side at the slurry side), in which filtration system the average pressure differential over the filtration medium is at least 0.1 bar, in which process the particle size distribution is such that at least a certain amount of the catalyst particles is smaller than the average pore size of the selective layer of the filtration medium. The invention also comprises an apparatus to carry out the process described above.

  12. Comparative assessment of TRU waste forms and processes. Volume I. Waste form and process evaluations

    SciTech Connect (OSTI)

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

    1982-09-01T23:59:59.000Z

    This study provides an assesses seven waste forms and eight processes for immobilizing transuranic (TRU) wastes. The waste forms considered are cast cement, cold-pressed cement, FUETAP (formed under elevated temperature and pressure) cement, borosilicate glass, aluminosilicate glass, basalt glass-ceramic, and cold-pressed and sintered silicate ceramic. The waste-immobilization processes considered are in-can glass melting, joule-heated glass melting, glass marble forming, cement casting, cement cold-pressing, FUETAP cement processing, ceramic cold-pressing and sintering, basalt glass-ceramic processing. Properties considered included gas generation, chemical durability, mechanical strength, thermal stability, and radiation stability. The ceramic products demonstrated the best properties, except for plutonium release during leaching. The glass and ceramic products had similar properties. The cement products generally had poorer properties than the other forms, except for plutonium release during leaching. Calculations of the Pu release indicated that the waste forms met the proposed NRC release rate limit of 1 part in 10/sup 5/ per year in most test conditions. The cast-cement process had the lowest processing cost, followed closely by the cold-pressed and FUETAP cement processes. Joule-heated glass melting had the lower cost of the glass processes. In-can melting in a high-quality canister had the highest cost, and cold-pressed and sintered ceramic the second highest. Labor and canister costs for in-can melting were identified. The major contributor to costs of disposing of TRU wastes in a defense waste repository is waste processing costs. Repository costs could become the dominant cost for disposing of TRU wastes in a commercial repository. It is recommended that cast and FUETAP cement and borosilicate glass waste-form systems be considered. 13 figures, 16 tables.

  13. Voluntary Protection Program Onsite, Liquid Waste Contract Savannah River Site- February 2011

    Broader source: Energy.gov [DOE]

    Evaluation to determine whether the Liquid Waste Contract Savannah River Site is continuing to perform at a level deserving DOE-VPP Star recognition.

  14. EA-1115: Liquid Waste Treatment at the Nevada Test Site, Nye County, Nevada

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of the proposal to treat low-level radioactive liquid and low-level mixed liquid and semi-solid wastes generated at the U.S. Department of Energy Nevada...

  15. Waste Minimization Study on Pyrochemical Reprocessing Processes

    SciTech Connect (OSTI)

    Boussier, H.; Conocar, O.; Lacquement, J. [CEA/DEN Valrho Marcoule/DRCP/SCPS/Pyrochemical Processes Laboratory, BP 17171 30207 Bagnols-sur-Ceze (France)

    2006-07-01T23:59:59.000Z

    Ideally a new pyro-process should not generate more waste, and should be at least as safe and cost effective as the hydrometallurgical processes currently implemented at industrial scale. This paper describes the thought process, the methodology and some results obtained by process integration studies to devise potential pyro-processes and to assess their capability of achieving this challenging objective. As example the assessment of a process based on salt/metal reductive extraction, designed for the reprocessing of Generation IV carbide spent fuels, is developed. Salt/metal reductive extraction uses the capability of some metals, aluminum in this case, to selectively reduce actinide fluorides previously dissolved in a fluoride salt bath. The reduced actinides enter the metal phase from which they are subsequently recovered; the fission products remain in the salt phase. In fact, the process is not so simple, as it requires upstream and downstream subsidiary steps. All these process steps generate secondary waste flows representing sources of actinide leakage and/or FP discharge. In aqueous processes the main solvent (nitric acid solution) has a low boiling point and evaporate easily or can be removed by distillation, thereby leaving limited flow containing the dissolved substance behind to be incorporated in a confinement matrix. From the point of view of waste generation, one main handicap of molten salt processes, is that the saline phase (fluoride in our case) used as solvent is of same nature than the solutes (radionuclides fluorides) and has a quite high boiling point. So it is not so easy, than it is with aqueous solutions, to separate solvent and solutes in order to confine only radioactive material and limit the final waste flows. Starting from the initial block diagram devised two years ago, the paper shows how process integration studies were able to propose process fittings which lead to a reduction of the waste variety and flows leading at an 'ideal' new block diagram allowing internal solvent recycling, and self eliminating reactants. This new flowsheet minimizes the quantity of inactive inlet flows that would have inevitably to be incorporated in a final waste form. The study identifies all knowledge gaps to be filled and suggest some possible R and D issues to confirm or infirm the feasibility of the proposed process fittings. (authors)

  16. Disposal of liquid radioactive wastes through wells or shafts

    SciTech Connect (OSTI)

    Perkins, B.L.

    1982-01-01T23:59:59.000Z

    This report describes disposal of liquids and, in some cases, suitable solids and/or entrapped gases, through: (1) well injection into deep permeable strata, bounded by impermeable layers; (2) grout injection into an impermeable host rock, forming fractures in which the waste solidifies; and (3) slurrying into excavated subsurface cavities. Radioactive materials are presently being disposed of worldwide using all three techniques. However, it would appear that if the techniques were verified as posing minimum hazards to the environment and suitable site-specific host rock were identified, these disposal techniques could be more widely used.

  17. Digestion of frozen/thawed food waste in the hybrid anaerobic solid-liquid system

    SciTech Connect (OSTI)

    Stabnikova, O. [School of Civil and Environmental Engineering, Nanyang Technological University, 50 Nanyang Avenue, Singapore 639798 (Singapore)], E-mail: costab@ntu.edu.sg; Liu, X.Y.; Wang, J.Y. [School of Civil and Environmental Engineering, Nanyang Technological University, 50 Nanyang Avenue, Singapore 639798 (Singapore)

    2008-07-01T23:59:59.000Z

    The hybrid anaerobic solid-liquid (HASL) system, which is a modified two-phase anaerobic digester, is to be used in an industrial scale operation to minimize disposal of food waste at incineration plants in Singapore. The aim of the present research was to evaluate freezing/thawing of food waste as a pre-treatment for its anaerobic digestion in the HASL system. The hydrolytic and fermentation processes in the acidogenic reactor were enhanced when food waste was frozen for 24 h at -20 deg. C and then thawed for 12 h at 25 deg. C (experiment) in comparison with fresh food waste (control). The highest dissolved COD concentrations in the leachate from the acidogenic reactors were 16.9 g/l on day 3 in the control and 18.9 g/l on day 1 in the experiment. The highest VFA concentrations in the leachate from the acidogenic reactors were 11.7 g/l on day 3 in the control and 17.0 g/l on day 1 in the experiment. The same volume of methane was produced during 12 days in the control and 7 days in the experiment. It gave the opportunity to diminish operational time of batch process by 42%. The effect of freezing/thawing of food waste as pre-treatment for its anaerobic digestion in the HASL system was comparable with that of thermal pre-treatment of food waste at 150 deg. C for 1 h. However, estimation of energy required either to heat the suspended food waste to 150 deg. C or to freeze the same quantity of food waste to -20 deg. C showed that freezing pre-treatment consumes about 3 times less energy than thermal pre-treatment.

  18. Defense Waste Processing Facility (DWPF), Modular CSSX Unit (CSSX), and Waste Transfer Line System of Salt Processing Program (U)

    SciTech Connect (OSTI)

    CHANG, ROBERT

    2006-02-02T23:59:59.000Z

    All of the waste streams from ARP, MCU, and SWPF processes will be sent to DWPF for vitrification. The impact these new waste streams will have on DWPF's ability to meet its canister production goal and its ability to support the Salt Processing Program (ARP, MCU, and SWPF) throughput needed to be evaluated. DWPF Engineering and Operations requested OBU Systems Engineering to evaluate DWPF operations and determine how the process could be optimized. The ultimate goal will be to evaluate all of the Liquid Radioactive Waste (LRW) System by developing process modules to cover all facilities/projects which are relevant to the LRW Program and to link the modules together to: (1) study the interfaces issues, (2) identify bottlenecks, and (3) determine the most cost effective way to eliminate them. The results from the evaluation can be used to assist DWPF in identifying improvement opportunities, to assist CBU in LRW strategic planning/tank space management, and to determine the project completion date for the Salt Processing Program.

  19. OVERVIEW OF TESTING TO SUPPORT PROCESSING OF SLUDGE BATCH 4 IN THE DEFENSE WASTE PROCESSING FACILITY

    SciTech Connect (OSTI)

    Herman, C

    2006-09-20T23:59:59.000Z

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site began processing of its third sludge batch in March 2004. To avoid a feed outage in the facility, the next sludge batch will have to be prepared and ready for transfer to the DWPF by the end of 2006. The next sludge batch, Sludge Batch 4 (SB4), will consist of a significant volume of HM-type sludge. HM-type sludge is very high in aluminum compared to the mostly Purex-type sludges that have been processed to date. The Savannah River National Laboratory (SRNL) has been working with Liquid Waste Operations to define the sludge preparation plans and to perform testing to support qualification and processing of SB4. Significant challenges have arisen during SB4 preparation and testing to include poor sludge settling behavior and lower than desired projected melt rates. An overview of the testing activities is provided.

  20. Improved FGD dewatering process cuts solid wastes

    SciTech Connect (OSTI)

    Moer, C.; Fernandez, J.; Carraro, B. [Duke Energy (United States)

    2009-08-15T23:59:59.000Z

    In 2007, Duke Energy's W.H. Zimmer Station set out to advance the overall performance of its flue gas desulfurization (FGD) dewatering process. The plant implemented a variety of measures, including upgrading water-solids separation, improving polymer program effectiveness and reliability, optimizing treatment costs, reducing solid waste sent to the landfill, decreasing labor requirements, and maintaining septic-free conditions in clarifiers. The changes succeeded in greatly reducing solid waste generation and achieving total annual savings of over half a million dollars per year. 8 figs., 1 tab.

  1. Waste minimization by process modification

    SciTech Connect (OSTI)

    Hopper, J.R.; Yaws, C.L.; Ho, T.C.; Vichailak, M. (Lamar Univ., Beaumont, TX (United States))

    1993-01-01T23:59:59.000Z

    A simulation of the Sohio process for the production of acrylonitrile from the catalytic ammoxidation of propylene has been performed, using published kinetic and thermodynamic data to illustrate the concepts of pollution prevention by process modification. The study has determined the reaction parameters which will minimize the production of by-products while maintaining the conversion of propylene above 80%. The reaction parameters studied were reactor type (plug flow reactor [PFR], continuous stirred tank reactor [CSTR], and fluidized bed reactor [FBC]), reaction temperature, residence time, and entering feed temperature. The minimum by-products were produced in an FBR operating at 450 C at a residence time of 7 seconds for a conversion of 81%.

  2. Process for treating alkaline wastes for vitrification

    DOE Patents [OSTI]

    Hsu, C.L.W.

    1995-07-25T23:59:59.000Z

    A process is described for treating alkaline wastes for vitrification. The process involves acidifying the wastes with an oxidizing agent such as nitric acid, then adding formic acid as a reducing agent, and then mixing with glass formers to produce a melter feed. The nitric acid contributes nitrates that act as an oxidant to balance the redox of the melter feed, prevent reduction of certain species to produce conducting metals, and lower the pH of the wastes to a suitable level for melter operation. The formic acid reduces mercury compounds to elemental mercury for removal by steam stripping, and MnO{sub 2} to the Mn(II) ion to prevent foaming of the glass melt. The optimum amounts of nitric acid and formic acid are determined in relation to the composition of the wastes, including the concentrations of mercury (II) and MnO{sub 2}, noble metal compounds, nitrates, formates and so forth. The process minimizes the amount of hydrogen generated during treatment, while producing a redox-balanced feed for effective melter operation and a quality glass product. 4 figs.

  3. Process for treating alkaline wastes for vitrification

    DOE Patents [OSTI]

    Hsu, Chia-lin W. (Augusta, GA)

    1995-01-01T23:59:59.000Z

    A process for treating alkaline wastes for vitrification. The process involves acidifying the wastes with an oxidizing agent such as nitric acid, then adding formic acid as a reducing agent, and then mixing with glass formers to produce a melter feed. The nitric acid contributes nitrates that act as an oxidant to balance the redox of the melter feed, prevent reduction of certain species to produce conducting metals, and lower the pH of the wastes to a suitable level for melter operation. The formic acid reduces mercury compounds to elemental mercury for removal by steam stripping, and MnO.sub.2 to the Mn(II) ion to prevent foaming of the glass melt. The optimum amounts of nitric acid and formic acid are determined in relation to the composition of the wastes, including the concentrations of mercury (II) and MnO.sub.2, noble metal compounds, nitrates, formates and so forth. The process minimizes the amount of hydrogen generated during treatment, while producing a redox-balanced feed for effective melter operation and a quality glass product.

  4. Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

    SciTech Connect (OSTI)

    Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

    1993-03-01T23:59:59.000Z

    A total of 6.05 x 10{sup 7} gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10{sup 6} gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system`s discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system`s discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10{sup 3} gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10{sup 4} gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10{sup 3} ft{sup 3}) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10{sup 2} ft{sup 3} of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10{sup 8} gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10{sup 7} gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10{sup 3} ft{sup 3}) of solid sludge generated by the precipitation/clarification process were removed from the filter press room.

  5. Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

    SciTech Connect (OSTI)

    Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

    1993-03-01T23:59:59.000Z

    A total of 6.05 x 10[sup 7] gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10[sup 6] gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system's discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system's discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10[sup 3] gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10[sup 4] gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10[sup 3] ft[sup 3]) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10[sup 2] ft[sup 3] of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10[sup 8] gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10[sup 7] gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10[sup 3] ft[sup 3]) of solid sludge generated by the precipitation/clarification process were removed from the filter press room.

  6. PROBABILITY BASED CORROSION CONTROL FOR LIQUID WASTE TANKS - PART III

    SciTech Connect (OSTI)

    Hoffman, E.; Edwards, T.

    2010-12-09T23:59:59.000Z

    The liquid waste chemistry control program is designed to reduce the pitting corrosion occurrence on tank walls. The chemistry control program has been implemented, in part, by applying engineering judgment safety factors to experimental data. However, the simple application of a general safety factor can result in use of excessive corrosion inhibiting agents. The required use of excess corrosion inhibitors can be costly for tank maintenance, waste processing, and in future tank closure. It is proposed that a probability-based approach can be used to quantify the risk associated with the chemistry control program. This approach can lead to the application of tank-specific chemistry control programs reducing overall costs associated with overly conservative use of inhibitor. Furthermore, when using nitrite as an inhibitor, the current chemistry control program is based on a linear model of increased aggressive species requiring increased protective species. This linear model was primarily supported by experimental data obtained from dilute solutions with nitrate concentrations less than 0.6 M, but is used to produce the current chemistry control program up to 1.0 M nitrate. Therefore, in the nitrate space between 0.6 and 1.0 M, the current control limit is based on assumptions that the linear model developed from data in the <0.6 M region is applicable in the 0.6-1.0 M region. Due to this assumption, further investigation of the nitrate region of 0.6 M to 1.0 M has potential for significant inhibitor reduction, while maintaining the same level of corrosion risk associated with the current chemistry control program. Ongoing studies have been conducted in FY07, FY08, FY09 and FY10 to evaluate the corrosion controls at the SRS tank farm and to assess the minimum nitrite concentrations to inhibit pitting in ASTM A537 carbon steel below 1.0 molar nitrate. The experimentation from FY08 suggested a non-linear model known as the mixture/amount model could be used to predict the probability of corrosion in ASTM A537 in varying solutions as shown in Figure 1. The mixture/amount model takes into account not only the ratio (or mixture) of inhibitors and aggressive species, but also the total concentration (or amount) of species in a solution. Historically, the ratio was the only factor taken into consideration in the development of the current chemistry control program. During FY09, an experimental program was undertaken to refine the mixture/amount model by further investigating the risk associated with reducing the minimum molar nitrite concentration required to confidently inhibit pitting in dilute solutions. The results of FY09, as shown in Figure 2, quantified the probability for a corrosion free outcome for combinations of nitrate and nitrite. The FY09 data predict probabilities up to 70%. Additional experimental data are needed to increase the probability to an acceptable percentage.

  7. Preliminary technical data summary No. 3 for the Defense Waste Processing Facility

    SciTech Connect (OSTI)

    Landon, L.F. (comp.)

    1980-05-01T23:59:59.000Z

    This document presents an update on the best information presently available for the purpose of establishing the basis for the design of a Defense Waste Processing Facility. Objective of this project is to provide a facility to fix the radionuclides present in Savannah River Plant (SRP) high-level liquid waste in a high-integrity form (glass). Flowsheets and material balances reflect the alternate CAB case including the incorporation of low-level supernate in concrete. (DLC)

  8. Rotor for processing liquids using movable capillary tubes

    DOE Patents [OSTI]

    Johnson, W.F.; Burtis, C.A.; Walker, W.A.

    1987-07-17T23:59:59.000Z

    A rotor assembly for processing liquids, especially whole blood samples, is disclosed. The assembly includes apparatus for separating non-liquid components of whole blood samples from liquid components, apparatus for diluting the separated liquid component with a diluent and apparatus for transferring the diluted sample to an external apparatus for analysis. The rotor assembly employs several movable capillary tubes to handle the sample and diluents. A method for using the rotor assembly to process liquids is also described. 5 figs.

  9. Rotor for processing liquids using movable capillary tubes

    DOE Patents [OSTI]

    Johnson, Wayne F. (Loudon, TN); Burtis, Carl A. (Oak Ridge, TN); Walker, William A. (Knoxville, TN)

    1989-05-30T23:59:59.000Z

    A rotor assembly for processing liquids, especially whole blood samples, is disclosed. The assembly includes apparatus for separating non-liquid components of whole blood samples from liquid components, apparatus for diluting the separated liquid component with a diluent and apparatus for transferring the diluted sample to an external apparatus for analysis. The rotor assembly employs several movable capillary tubes to handle the sample and diluents. A method for using the rotor assembly to process liquids is also described.

  10. Rotor for processing liquids using movable capillary tubes

    DOE Patents [OSTI]

    Johnson, Wayne F. (Loudon, TN); Burtis, Carl A. (Oak Ridge, TN); Walker, William A. (Knoxville, TN)

    1989-01-01T23:59:59.000Z

    A rotor assembly for processing liquids, especially whole blood samples, is disclosed. The assembly includes apparatus for separating non-liquid components of whole blood samples from liquid components, apparatus for diluting the separated liquid component with a diluent and apparatus for transferring the diluted sample to an external apparatus for analysis. The rotor assembly employs several movable capillary tubes to handle the sample and diluents. A method for using the rotor assembly to process liquids is also described.

  11. Crystallization during processing of nuclear waste glass

    SciTech Connect (OSTI)

    Hrma, Pavel R.

    2010-12-01T23:59:59.000Z

    In glass processing situations involving glass crystallization, various crystalline forms nucleate, grow, and dissolve, typically in a nonuniform temperature field of molten glass subjected to convection. Nuclear waste glasses are remarkable examples of multicomponent vitrified mixtures involving partial crystallization. In the glass melter, crystals form and dissolve during batch-to-glass conversion, melter processing, and product cooling. Crystals often agglomerate and sink, and they may settle at the melter bottom. Within the body of cooling glass, multiple phases crystallize in a non-uniform time-dependent temperature field. Self-organizing periodic distribution (the Liesegnang effect) is common. Various crystallization phenomena that occur in glassmaking are reviewed.

  12. Independent Oversight Review, Oak Ridge Transuranic Waste Processing...

    Office of Environmental Management (EM)

    of Safety Systems at the Oak Ridge Transuranic Waste Processing Center and Associated Feedback and Improvement Processes. This report documents the results of an independent...

  13. Waste Heat Management Options for Improving Industrial Process...

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

    Heat Management Options for Improving Industrial Process Heating Systems Waste Heat Management Options for Improving Industrial Process Heating Systems This presentation covers...

  14. INSTALLATION OF BUBBLERS IN THE SAVANNAH RIVER SITED DEFENSE WASTE PROCESSING FACILITY MELTER

    SciTech Connect (OSTI)

    Smith, M.; Iverson, D.

    2010-12-08T23:59:59.000Z

    Savannah River Remediation (SRR) LLC assumed the liquid waste contract at the Savannah River Site (SRS) in the summer of 2009. The main contractual agreement was to close 22 High Level Waste (HLW) tanks in eight years. To achieve this aggressive commitment, faster waste processing throughout the SRS liquid waste facilities will be required. Part of the approach to achieve faster waste processing is to increase the canister production rate of the Defense Waste Processing Facility (DWPF) from approximately 200 canisters filled with radioactive waste glass per year to 400 canisters per year. To reach this rate for melter throughput, four bubblers were installed in the DWPF Melter in the late summer of 2010. This effort required collaboration between SRR, SRR critical subcontractor EnergySolutions, and Savannah River Nuclear Solutions, including the Savannah River National Laboratory (SRNL). The tasks included design and fabrication of the bubblers and related equipment, testing of the bubblers for various technical issues, the actual installation of the bubblers and related equipment, and the initial successful operation of the bubblers in the DWPF Melter.

  15. Double Shell Tank (DST) Process Waste Sampling Subsystem Specification

    SciTech Connect (OSTI)

    RASMUSSEN, J.H.

    2000-05-03T23:59:59.000Z

    This specification establishes the performance requirements and provides references to the requisite codes and standards to be applied to the Double-Shell Tank (DST) Process Waste Sampling Subsystem which supports the first phase of Waste Feed Delivery.

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

    SciTech Connect (OSTI)

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

    2013-08-22T23:59:59.000Z

    The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be produced from a melting and crystallization process. The objective of this report is to explore the phase formation and microstructural differences between lab scale melt processing in varying gas environments with alternative densification processes such as Hot Pressing (HP) and Spark Plasma Sintering (SPS). The waste stream used as the basis for the development and testing is a simulant derived from a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. Melt processing as well as solid state sintering routes SPS and HP demonstrated the formation of the targeted phases; however differences in microstructure and elemental partitioning were observed. In SPS and HP samples, hollandite, pervoskite/pyrochlore, zirconolite, metallic alloy and TiO{sub 2} and Al{sub 2}O{sub 3} were observed distributed in a network of fine grains with small residual pores. The titanate phases that incorporate M{sup +3} rare earth elements were observed to be distinct phases (ex. Nd{sub 2}Ti{sub 2}O{sub 7}) with less degree of substitution as compared to the more homogeneous melt processed samples where a high degree of substitution and variation of composition within grains was observed. Liquid phase sintering was enhanced in reducing gas environments and resulted in large (10-200 microns) irregular shaped grains along with large voids associated with the melt process; SPS and HP samples exhibited finer grain size with smaller voids. Metallic alloys were observed in the bulk of the sample for SPS and HP samples, but were found at the bottom of the crucible in melt processed trials. These results indicate that for a first melter trial, the targeted phases can be formed in air by utilizing Ti/TiO{sub 2} additives which aid phase formation and improve the electrical conductivity. Ultimately, a melter run in reducing gas environments would be beneficial to study differences in phase formation and elemental partitioning.

  17. GEOTECHNICAL/GEOCHEMICAL CHARACTERIZATION OF ADVANCED COAL PROCESS WASTE STREAMS

    SciTech Connect (OSTI)

    Edwin S. Olson; Charles J. Moretti

    1999-11-01T23:59:59.000Z

    Thirteen solid wastes, six coals and one unreacted sorbent produced from seven advanced coal utilization processes were characterized for task three of this project. The advanced processes from which samples were obtained included a gas-reburning sorbent injection process, a pressurized fluidized-bed coal combustion process, a coal-reburning process, a SO{sub x}, NO{sub x}, RO{sub x}, BOX process, an advanced flue desulfurization process, and an advanced coal cleaning process. The waste samples ranged from coarse materials, such as bottom ashes and spent bed materials, to fine materials such as fly ashes and cyclone ashes. Based on the results of the waste characterizations, an analysis of appropriate waste management practices for the advanced process wastes was done. The analysis indicated that using conventional waste management technology should be possible for disposal of all the advanced process wastes studied for task three. However, some wastes did possess properties that could present special problems for conventional waste management systems. Several task three wastes were self-hardening materials and one was self-heating. Self-hardening is caused by cementitious and pozzolanic reactions that occur when water is added to the waste. All of the self-hardening wastes setup slowly (in a matter of hours or days rather than minutes). Thus these wastes can still be handled with conventional management systems if care is taken not to allow them to setup in storage bins or transport vehicles. Waste self-heating is caused by the exothermic hydration of lime when the waste is mixed with conditioning water. If enough lime is present, the temperature of the waste will rise until steam is produced. It is recommended that self-heating wastes be conditioned in a controlled manner so that the heat will be safely dissipated before the material is transported to an ultimate disposal site. Waste utilization is important because an advanced process waste will not require ultimate disposal when it is put to use. Each task three waste was evaluated for utilization potential based on its physical properties, bulk chemical composition, and mineral composition. Only one of the thirteen materials studied might be suitable for use as a pozzolanic concrete additive. However, many wastes appeared to be suitable for other high-volume uses such as blasting grit, fine aggregate for asphalt concrete, road deicer, structural fill material, soil stabilization additives, waste stabilization additives, landfill cover material, and pavement base course construction.

  18. active liquid wastes: Topics by E-print Network

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

    Assessment Institute Rethinking the Waste Hierarchy March 2005 Recommendations A number of specific recommendations for achieving cost-effective waste policies can be made...

  19. active liquid waste: Topics by E-print Network

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

    Assessment Institute Rethinking the Waste Hierarchy March 2005 Recommendations A number of specific recommendations for achieving cost-effective waste policies can be made...

  20. Buried waste integrated demonstration technology integration process

    SciTech Connect (OSTI)

    Ferguson, J.S.; Ferguson, J.E.

    1992-04-01T23:59:59.000Z

    A Technology integration Process was developed for the Idaho National Energy Laboratories (INEL) Buried Waste Integrated Demonstration (BWID) Program to facilitate the transfer of technology and knowledge from industry, universities, and other Federal agencies into the BWID; to successfully transfer demonstrated technology and knowledge from the BWID to industry, universities, and other Federal agencies; and to share demonstrated technologies and knowledge between Integrated Demonstrations and other Department of Energy (DOE) spread throughout the DOE Complex. This document also details specific methods and tools for integrating and transferring technologies into or out of the BWID program. The document provides background on the BWID program and technology development needs, demonstrates the direction of technology transfer, illustrates current processes for this transfer, and lists points of contact for prospective participants in the BWID technology transfer efforts. The Technology Integration Process was prepared to ensure compliance with the requirements of DOE's Office of Technology Development (OTD).

  1. Buried waste integrated demonstration technology integration process

    SciTech Connect (OSTI)

    Ferguson, J.S.; Ferguson, J.E.

    1992-04-01T23:59:59.000Z

    A Technology integration Process was developed for the Idaho National Energy Laboratories (INEL) Buried Waste Integrated Demonstration (BWID) Program to facilitate the transfer of technology and knowledge from industry, universities, and other Federal agencies into the BWID; to successfully transfer demonstrated technology and knowledge from the BWID to industry, universities, and other Federal agencies; and to share demonstrated technologies and knowledge between Integrated Demonstrations and other Department of Energy (DOE) spread throughout the DOE Complex. This document also details specific methods and tools for integrating and transferring technologies into or out of the BWID program. The document provides background on the BWID program and technology development needs, demonstrates the direction of technology transfer, illustrates current processes for this transfer, and lists points of contact for prospective participants in the BWID technology transfer efforts. The Technology Integration Process was prepared to ensure compliance with the requirements of DOE`s Office of Technology Development (OTD).

  2. The Mochovce final treatment center for liquid radioactive waste introduced to active trial operation

    SciTech Connect (OSTI)

    Krajc, T.; Stubna, M.; Kravarik, K.; Zatkulak, M. [VUJE Trnava, Inc. (Slovakia); Slezak, M.; Remias, V. [Javys - Jadrova a vyradovacia spolocnost, a.s. - Nuclear and Decommissioning Company, plc., Tomasikova 22, 821 02 Bratislava (Slovakia)

    2007-07-01T23:59:59.000Z

    The Final Treatment Centre (FTC) for Mochovce Nuclear Power Plant (NPP) have been designed for treatment and final conditioning of radioactive liquid and wet waste produced by named NPP equipped with Russian VVER-440 type of reactors. Treated wastes comprise radioactive concentrates, spent resin and sludge. VUJE Inc. as an experienced company in field of treatment of radioactive waste in Slovakia has been chosen as main contractor for technological part of FTC. During the realisation of project the future operator of Centre required the contractor to solve the treatment of wastes produced in the process of NPP A-1 decommissioning. On the basis of this requirement the project was modified in order to enable manipulations with waste products from A-1 NPP transported to Centre in steel drums. The initial project was prepared in 2003. The design and manufacture of main components were performed in 2004 and 2005. FTC civil works started in August 2004. Initial nonradioactive testing of the system parts were carried out from April to September 2006, then the tests of systems started with model concentrates and non-radioactive resins. After the processes evaluation the radioactive test performed from February 2007. A one-year trial operation of facility is planned for completion during 2007 and 2008. The company JAVYS, Inc. is responsible for radioactive waste and spent fuel treatment in the Slovak republic and will operate the FTC during trial operation and after its completion. This Company has also significant experience with operation of Jaslovske Bohunice Treatment Centre. The overall capacity of the FTC is 820 m{sup 3}/year of concentrates and 40 m{sup 3}/year of spent resin and sludge. Bituminization and cementation were provided as main technologies for treatment of these wastes. Treatment of concentrate is performed by bituminization on Thin Film Evaporator with rotating wiping blades. Spent resin and sludge are decanted, dried and mixed with bitumen in blade homogeniser. The bitumen product is discharged into 200 dm{sup 3} steel drums. Drums with bitumen product or drums originated from A-1 NPP are loaded into Fibre Reinforced Concrete containers (FRC) and grouted with cement. Cement grout is prepared from the mixture of cement, additive and radioactive over-concentrate. By formulating the cement grout with evaporator concentrates the maximum radioactivity is fixed in cement matrix and volume of final waste product is minimized. A batch mixer with rotating blades is used to produce the cement grout. The grouted FRC containers are stored in the expedition hall and after 28 days of curing are transported to final disposal. After the start of routine operation, the FTC provides treatment for all liquid and wet LLW produced from the operation of the Mochovce NPP. The final product of the FTC is a FRC loaded with bitumen product in drums and filled with radioactive cement product. This container meets all limits for final disposal in the National Radioactive Waste Repository at Mochovce. This paper introducing the main parts of FTC and describes the technological procedures including the basic technological parameters for both used technologies, their working capacity and the overall waste flow. The evaluation of experience gained in the phases of Centre construction and commissioning and partially trial operation as well is a part of this paper (Evaluation of completion works process and time schedule, the process of individual system parts testing, testing of systems using model media, radioactive testing and trial operation). (authors)

  3. Processing constraints on high-level nuclear waste glasses for Hanford Waste Vitrification Plant

    SciTech Connect (OSTI)

    Hrma, P. [Pacific Northwest Lab., Richland, WA (United States)

    1993-12-31T23:59:59.000Z

    The work presented in this paper is a part of a major technology program supported by the US Department of Energy (DOE) in preparation for the planned operation of the Hanford Waste Vitrification Plant (HWVP). Because composition of Hanford waste varies greatly, processability is a major concern for successful vitrification. This paper briefly surveys general aspects of waste glass processability and then discusses their ramifications for specific examples of Hanford waste streams.

  4. Studien-und Prfungsordnung der Universitt Stuttgart fr den auslandsorientierten Studiengang Air Quality Control, Solid Waste and Waste Water Process Engineering

    E-Print Network [OSTI]

    Reyle, Uwe

    Air Quality Control, Solid Waste and Waste Water Process Engineering (WASTE) mit Abschluss Master Quality Control, Solid Waste and Waste Water Process Engineering" (WASTE) beschlossen. Der Rektor hat Control, Solid Waste and Waste Water Process Engineering" (WASTE) überblickt werden, die Fähigkeit

  5. Industrial waste reduction: The process problem

    SciTech Connect (OSTI)

    Valentino, F.W.; Walmet, G.E.

    1986-09-01T23:59:59.000Z

    Industrial waste problems, especially those involving hazardous waste, seem to be pervasive. The national media report newly discovered waste problems and sites with alarming regularity. Examples that immediately come to mind are Love Canal, New York; Times Beach, Missouri; and Seveso, Italy. Public perceptions of the industrial waste problem, reflecting the media's focus, appear to be that: large corporations are solely responsible for creating waste dumps, and the only role of government is to prevent illegal dumping and to regulate, fine, and require corporations to rectify the problem; all efforts should be directed toward preventing illegal dumping and treatment of the existing waste dumps; all industrial wastes can be classified as hazardous in nature. This general impression is both inaccurate and incomplete. All industrial waste is not hazardous (although most of it is not benign). All waste producers are not large corporations: nearly all industries produce some wastes. And, while existing waste sites must be effectively treated, additional efforts are needed at other points in the industrial waste cycle. Most people would agree both that waste dumping must be carefully regulated because of its negative impacts on the environment and that the less waste the better, even with carefully regulated disposal. Since nearly all industry now produces some waste and no one expects industry to shut down to resolve the waste problem, other strategies need to be available to deal with the problem at the front end. This paper discusses alternative strategies.

  6. Liquid and Gaseous Waste Operations Department annual operating report, CY 1995

    SciTech Connect (OSTI)

    Maddox, J.J.; Scott, C.B.

    1996-03-01T23:59:59.000Z

    This report describes the operating activities, upgrade activities, maintenance, and other activities regarding liquid and gaseous low level radioactive waste management at the Oak Ridge National Laboratory. Miscellaneous activities include training, audits, tours, and environmental restoration support.

  7. High-Level waste process and product data annotated bibliography

    SciTech Connect (OSTI)

    Stegen, G.E.

    1996-02-13T23:59:59.000Z

    The objective of this document is to provide information on available issued documents that will assist interested parties in finding available data on high-level waste and transuranic waste feed compositions, properties, behavior in candidate processing operations, and behavior on candidate product glasses made from those wastes. This initial compilation is only a partial list of available references.

  8. 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.

  9. Waste receiving and processing facility module 1 auditable safetyanalysis

    SciTech Connect (OSTI)

    Bottenus, R.J.

    1997-02-01T23:59:59.000Z

    The Waste Receiving and Processing Facility Module 1 Auditable Safety Analysis analyzes postulated accidents and determines controls to prevent the accidents or mitigate the consequences.

  10. Transuranic Waste Processing Center Oak Ridge Site Specific...

    Office of Environmental Management (EM)

    Transuranic Waste Processing Update Oak Ridge Site Specific Advisory Board May 14, 2014 Laura Wilkerson, Portfolio Federal Project Director Karen Deacon, Deputy Federal Project...

  11. Transuranic Waste Processing Center (TWPC) Legacy Tank RH-TRU Sludge Processing and Compliance Strategy - 13255

    SciTech Connect (OSTI)

    Rogers, Ben C.; Heacker, Fred K.; Shannon, Christopher [Wastren Advantage, Inc., Transuranic Waste Processing Center, 100 WIPP Road, Lenoir City, Tennessee 37771 (United States)] [Wastren Advantage, Inc., Transuranic Waste Processing Center, 100 WIPP Road, Lenoir City, Tennessee 37771 (United States); and others

    2013-07-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) needs to safely and efficiently treat its 'legacy' transuranic (TRU) waste and mixed low-level waste (LLW) from past research and defense activities at the Oak Ridge National Laboratory (ORNL) so that the waste is prepared for safe and secure disposal. The TWPC operates an Environmental Management (EM) waste processing facility on the Oak Ridge Reservation (ORR). The TWPC is classified as a Hazard Category 2, non-reactor nuclear facility. This facility receives, treats, and packages low-level waste and TRU waste stored at various facilities on the ORR for eventual off-site disposal at various DOE sites and commercial facilities. The Remote Handled TRU Waste Sludge held in the Melton Valley Storage Tanks (MVSTs) was produced as a result of the collection, treatment, and storage of liquid radioactive waste originating from the ORNL radiochemical processing and radioisotope production programs. The MVSTs contain most of the associated waste from the Gunite and Associated Tanks (GAAT) in the ORNL's Tank Farms in Bethel Valley and the sludge (SL) and associated waste from the Old Hydro-fracture Facility tanks and other Federal Facility Agreement (FFA) tanks. The SL Processing Facility Build-outs (SL-PFB) Project is integral to the EM cleanup mission at ORNL and is being accelerated by DOE to meet updated regulatory commitments in the Site Treatment Plan. To meet these commitments a Baseline (BL) Change Proposal (BCP) is being submitted to provide continued spending authority as the project re-initiation extends across fiscal year 2012 (FY2012) into fiscal year 2013. Future waste from the ORNL Building 3019 U-233 Disposition project, in the form of U-233 dissolved in nitric acid and water, down-blended with depleted uranyl nitrate solution is also expected to be transferred to the 7856 MVST Annex Facility (formally the Capacity Increase Project (CIP) Tanks) for co-processing with the SL. The SL-PFB project will construct and install the necessary integrated systems to process the accumulated MVST Facilities SL inventory at the TWPC thus enabling safe and effective disposal of the waste. This BCP does not include work to support current MVST Facility Surveillance and Maintenance programs or the ORNL Building 3019 U-233 Disposition project, since they are not currently part of the TWPC prime contract. The purpose of the environmental compliance strategy is to identify the environmental permits and other required regulatory documents necessary for the construction and operation of the SL- PFB at the TWPC, Oak Ridge, TN. The permits and other regulatory documents identified are necessary to comply with the environmental laws and regulations of DOE Orders, and other requirements documented in the SL-PFB, Safety Design Strategy (SDS), SL-A-AD-002, R0 draft, and the Systems, Function and Requirements Document (SFRD), SL-X-AD-002, R1 draft. This compliance strategy is considered a 'living strategy' and it is anticipated that it will be revised as design progresses and more detail is known. The design basis on which this environmental permitting and compliance strategy is based is the Wastren Advantage, Inc., (WAI), TWPC, SL-PFB (WAI-BL-B.01.06) baseline. (authors)

  12. SUCCESSES AND EMERGING ISSUES IN SIMULATING THE MIXING BEHAVIOR OF LIQUID-PARTICLE NUCLEAR WASTE SLURRIES AT THE SAVANNAH RIVER SITE - 211B

    SciTech Connect (OSTI)

    Koopman, D.; Pickenheim, B.; Lambert, D.; Stone, M.

    2009-09-02T23:59:59.000Z

    Aqueous radioactive high-level waste slurries are combined during processing steps that ultimately produce a stable borosilicate glass waste form. Chemically treated waste slurries are combined with each other and with glass frit-water slurries to produce the melter feed. Understanding the evolution of the rheological properties of the slurries is an important aspect of removing and treating the stored waste. To a first approximation, combinations of colloidal waste slurry with {approx}0.1-mm mean diameter glass frit or glass beads act in an analogous matter to slurries of spherical beads in Newtonian liquids. The non-Newtonian rheological properties of the waste slurries without frit, however, add complexity to the hydrodynamic analysis. The use of shear rate dependent apparent viscosities with the modified Einstein equation was used to model the rheological properties of aqueous frit-waste slurries.

  13. Process for treating fission waste. [Patent application

    DOE Patents [OSTI]

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

    1981-11-17T23:59:59.000Z

    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.

  14. Waste Analysis Plan for the Waste Receiving and Processing (WRAP) Facility

    SciTech Connect (OSTI)

    TRINER, G.C.

    1999-11-01T23:59:59.000Z

    The purpose of this waste analysis plan (WAP) is to document the waste acceptance process, sampling methodologies, analytical techniques, and overall processes that are undertaken for dangerous, mixed, and radioactive waste accepted for confirmation, nondestructive examination (NDE) and nondestructive assay (NDA), repackaging, certification, and/or storage at the Waste Receiving and Processing Facility (WRAP). Mixed and/or radioactive waste is treated at WRAP. WRAP is located in the 200 West Area of the Hanford Facility, Richland, Washington. Because dangerous waste does not include source, special nuclear, and by-product material components of mixed waste, radionuclides are not within the scope of this documentation. The information on radionuclides is provided only for general knowledge.

  15. Hydrothermal processing of Hanford tank wastes: Process modeling and control

    SciTech Connect (OSTI)

    Currier, R.P. [comp.

    1994-10-01T23:59:59.000Z

    In the Los Alamos National Laboratory (LANL) hydrothermal process, waste streams are first pressurized and heated as they pass through a continuous flow tubular reactor vessel. The waste is maintained at reaction temperature of 300--550 C where organic destruction and sludge reformation occur. This report documents LANL activities in process modeling and control undertaken in FY94 to support hydrothermal process development. Key issues discussed include non-ideal flow patterns (e.g. axial dispersion) and their effect on reactor performance, the use and interpretation of inert tracer experiments, and the use of computational fluid mechanics to evaluate novel hydrothermal reactor designs. In addition, the effects of axial dispersion (and simplifications to rate expressions) on the estimated kinetic parameters are explored by non-linear regression to experimental data. Safety-related calculations are reported which estimate the explosion limits of effluent gases and the fate of hydrogen as it passes through the reactor. Development and numerical solution of a generalized one-dimensional mathematical model is also summarized. The difficulties encountered in using commercially available software to correlate the behavior of high temperature, high pressure aqueous electrolyte mixtures are summarized. Finally, details of the control system and experiments conducted to empirically determine the system response are reported.

  16. Thermal processing system concepts and considerations for RWMC buried waste

    SciTech Connect (OSTI)

    Eddy, T.L.; Kong, P.C.; Raivo, B.D.; Anderson, G.L.

    1992-02-01T23:59:59.000Z

    This report presents a preliminary determination of ex situ thermal processing system concepts and related processing considerations for application to remediation of transuranic (TRU)-contaminated buried wastes (TRUW) at the Radioactive Waste Management Complex (RWMC) of the Idaho National Engineering Laboratory (INEL). Beginning with top-level thermal treatment concepts and requirements identified in a previous Preliminary Systems Design Study (SDS), a more detailed consideration of the waste materials thermal processing problem is provided. Anticipated waste stream elements and problem characteristics are identified and considered. Final waste form performance criteria, requirements, and options are examined within the context of providing a high-integrity, low-leachability glass/ceramic, final waste form material. Thermal processing conditions required and capability of key systems components (equipment) to provide these material process conditions are considered. Information from closely related companion study reports on melter technology development needs assessment and INEL Iron-Enriched Basalt (IEB) research are considered. Five potentially practicable thermal process system design configuration concepts are defined and compared. A scenario for thermal processing of a mixed waste and soils stream with essentially no complex presorting and using a series process of incineration and high temperature melting is recommended. Recommendations for applied research and development necessary to further detail and demonstrate the final waste form, required thermal processes, and melter process equipment are provided.

  17. Biological Information Document, Radioactive Liquid Waste Treatment Facility

    SciTech Connect (OSTI)

    Biggs, J.

    1995-12-31T23:59:59.000Z

    This document is intended to act as a baseline source material for risk assessments which can be used in Environmental Assessments and Environmental Impact Statements. The current Radioactive Liquid Waste Treatment Facility (RLWTF) does not meet current General Design Criteria for Non-reactor Nuclear Facilities and could be shut down affecting several DOE programs. This Biological Information Document summarizes various biological studies that have been conducted in the vicinity of new Proposed RLWTF site and an Alternative site. The Proposed site is located on Mesita del Buey, a mess top, and the Alternative site is located in Mortandad Canyon. The Proposed Site is devoid of overstory species due to previous disturbance and is dominated by a mixture of grasses, forbs, and scattered low-growing shrubs. Vegetation immediately adjacent to the site is a pinyon-juniper woodland. The Mortandad canyon bottom overstory is dominated by ponderosa pine, willow, and rush. The south-facing slope was dominated by ponderosa pine, mountain mahogany, oak, and muhly. The north-facing slope is dominated by Douglas fir, ponderosa pine, and oak. Studies on wildlife species are limited in the vicinity of the proposed project and further studies will be necessary to accurately identify wildlife populations and to what extent they utilize the project area. Some information is provided on invertebrates, amphibians and reptiles, and small mammals. Additional species information from other nearby locations is discussed in detail. Habitat requirements exist in the project area for one federally threatened wildlife species, the peregrine falcon, and one federal candidate species, the spotted bat. However, based on surveys outside of the project area but in similar habitats, these species are not expected to occur in either the Proposed or Alternative RLWTF sites. Habitat Evaluation Procedures were used to evaluate ecological functioning in the project area.

  18. Oak Ridge National Lebroatory Liquid&Gaseous Waste Treatment System Strategic Plan

    SciTech Connect (OSTI)

    Van Hoesen, S.D.

    2003-09-09T23:59:59.000Z

    Excellence in Laboratory operations is one of the three key goals of the Oak Ridge National Laboratory (ORNL) Agenda. That goal will be met through comprehensive upgrades of facilities and operational approaches over the next few years. Many of ORNL's physical facilities, including the liquid and gaseous waste collection and treatment systems, are quite old, and are reaching the end of their safe operating life. The condition of research facilities and supporting infrastructure, including the waste handling facilities, is a key environmental, safety and health (ES&H) concern. The existing infrastructure will add considerably to the overhead costs of research due to increased maintenance and operating costs as these facilities continue to age. The Liquid Gaseous Waste Treatment System (LGWTS) Reengineering Project is a UT-Battelle, LLC (UT-B) Operations Improvement Program (OIP) project that was undertaken to develop a plan for upgrading the ORNL liquid and gaseous waste systems to support ORNL's research mission.

  19. Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

    SciTech Connect (OSTI)

    Digby Macdonald; Brian Marx; Balaji Soundararajan; Morgan Smith

    2005-07-28T23:59:59.000Z

    The different tasks that have been carried out under the current program are as follows: (1) Theoretical and experimental assessment of general corrosion of iron/steel in borate buffer solutions by using electrochemical impedance spectroscopy (EIS), ellipsometry and XPS techniques; (2) Development of a damage function analysis (DFA), which would help in predicting the accumulation of damage due to pitting corrosion in an environment prototypical of DOE liquid waste systems; (3) Experimental measurement of crack growth rate, acoustic emission signals, and coupling currents for fracture in carbon and low alloy steels as functions of mechanical (stress intensity), chemical (conductivity), electrochemical (corrosion potential, ECP), and microstructural (grain size, precipitate size, etc) variables in a systematic manner, with particular attention being focused on the structure of the noise in the current and its correlation with the acoustic emissions; (4) Development of fracture mechanisms for carbon and low alloy steels that are consistent with the crack growth rate, coupling current data and acoustic emissions; (5) Inserting advanced crack growth rate models for SCC into existing deterministic codes for predicting the evolution of corrosion damage in DOE liquid waste storage tanks; (6) Computer simulation of the anodic and cathodic activity on the surface of the steel samples in order to exactly predict the corrosion mechanisms; (7) Wavelet analysis of EC noise data from steel samples undergoing corrosion in an environment similar to that of the high level waste storage containers, to extract data pertaining to general, pitting and stress corrosion processes, from the overall data. The work has yielded a number of important findings, including an unequivocal demonstration of the role of chloride ion in passivity breakdown on nickel in terms of cation vacancy generation within the passive film, the first detection and characterization of individual micro fracture events in stress corrosion cracking, and the determination of kinetic parameters for the generation and annihilation of point defects in the passive film on iron. The existence of coupling between the internal crack environment and the external cathodic environment, as predicted by the coupled environment fracture model (CEFM), has also been indisputably established for the AISI 4340/NaOH system. It is evident from the studies that analysis of coupling current noise is a very sensitive tool for studying the crack tip processes in relation to the chemical, mechanical, electrochemical, and microstructural properties of the system. Experiments are currently being carried out to explore these crack tip processes by simultaneous measurement of the acoustic activity at the crack tip in an effort to validate the coupling current data. These latter data are now being used to deterministically predict the accumulation of general and localized corrosion damage on carbon in prototypical DOE liquid waste storage tanks. Computer simulation of the cathodic and anodic activity on the steel surfaces is also being carried out in an effort to simulate the actual corrosion process. Wavelet analysis of the coupling current data promises to be a useful tool to differentiate between the different corrosion mechanisms. Hence, wavelet analysis of the coupling current data from the DOE waste containers is also being carried out to extract data pertaining to general, pitting and stress corrosion processes, from the overall data which is bound to contain noise fluctuations due to any or all of the above mentioned processes.

  20. System for removing liquid waste from a tank

    DOE Patents [OSTI]

    Meneely, Timothy K. (Penn Hills, PA); Sherbine, Catherine A. (N. Versailles Township, Allegheny County, PA)

    1994-01-01T23:59:59.000Z

    A tank especially suited for nuclear applications is disclosed. The tank comprises a tank shell for protectively surrounding the liquid contained therein; an inlet positioned on the tank for passing a liquid into the tank; a sump positioned in an interior portion of the tank for forming a reservoir of the liquid; a sloped incline for resting the tank thereon and for creating a natural flow of the liquid toward the sump; a pump disposed adjacent the tank for pumping the liquid; and a pipe attached to the pump and extending into the sump for passing the liquid therethrough. The pump pumps the liquid in the sump through the pipe and into the pump for discharging the liquid out of the tank.

  1. System for removing liquid waste from a tank

    DOE Patents [OSTI]

    Meneely, T.K.; Sherbine, C.A.

    1994-04-26T23:59:59.000Z

    A tank especially suited for nuclear applications is disclosed. The tank comprises a tank shell for protectively surrounding the liquid contained therein; an inlet positioned on the tank for passing a liquid into the tank; a sump positioned in an interior portion of the tank for forming a reservoir of the liquid; a sloped incline for resting the tank thereon and for creating a natural flow of the liquid toward the sump; a pump disposed adjacent the tank for pumping the liquid; and a pipe attached to the pump and extending into the sump for passing the liquid there through. The pump pumps the liquid in the sump through the pipe and into the pump for discharging the liquid out of the tank. 2 figures.

  2. Tank waste remediation system phase I high-level waste feed processability assessment report

    SciTech Connect (OSTI)

    Lambert, S.L.; Stegen, G.E., Westinghouse Hanford

    1996-08-01T23:59:59.000Z

    This report evaluates the effects of feed composition on the Phase I high-level waste immobilization process and interim storage facility requirements for the high-level waste glass.Several different Phase I staging (retrieval, blending, and pretreatment) scenarios were used to generate example feed compositions for glass formulations, testing, and glass sensitivity analysis. Glass models and data form laboratory glass studies were used to estimate achievable waste loading and corresponding glass volumes for various Phase I feeds. Key issues related to feed process ability, feed composition, uncertainty, and immobilization process technology are identified for future consideration in other tank waste disposal program activities.

  3. Foaming and Antifoaming in Radioactive Waste Pretreatment and Immobilization Processes

    SciTech Connect (OSTI)

    Darsh T. Wasan; Alex D. Nikolov; D.P. Lamber; T. Bond Calloway; M.E. Stone

    2005-03-12T23:59:59.000Z

    Savannah River National Laboratory (SRNL) has reported severe foaminess in the bench scale evaporation of the Hanford River Protection - Waste Treatment Plant (RPP-WPT) envelope C waste. Excessive foaming in waste evaporators can cause carryover of radionuclides and non-radioactive waste to the condensate system. The antifoams used at Hanford and tested by SRNL are believed to degrade and become inactive in high pH solutions. Hanford wastes have been known to foam during evaporation causing excessive down time and processing delays.

  4. Final Report - "Foaming and Antifoaming and Gas Entrainment in Radioactive Waste Pretreatment and Immobilization Processes"

    SciTech Connect (OSTI)

    Wasan, Darsh T.

    2007-10-09T23:59:59.000Z

    The Savannah River Site (SRS) and Hanford site are in the process of stabilizing millions of gallons of radioactive waste slurries remaining from production of nuclear materials for the Department of Energy (DOE). The Defense Waste Processing Facility (DWPF) at SRS is currently vitrifying the waste in borosilicate glass, while the facilities at the Hanford site are in the construction phase. Both processes utilize slurry-fed joule-heated melters to vitrify the waste slurries. The DWPF has experienced difficulty during operations. The cause of the operational problems has been attributed to foaming, gas entrainment and the rheological properties of the process slurries. The rheological properties of the waste slurries limit the total solids content that can be processed by the remote equipment during the pretreatment and meter feed processes. Highly viscous material can lead to air entrainment during agitation and difficulties with pump operations. Excessive foaming in waste evaporators can cause carryover of radionuclides and non-radioactive waste to the condensate system. Experimental and theoretical investigations of the surface phenomena, suspension rheology and bubble generation of interactions that lead to foaming and air entrainment problems in the DOE High Level and Low Activity Radioactive Waste separation and immobilization processes were pursued under this project. The first major task accomplished in the grant proposal involved development of a theoretical model of the phenomenon of foaming in a three-phase gas-liquid-solid slurry system. This work was presented in a recently completed Ph.D. thesis (9). The second major task involved the investigation of the inter-particle interaction and microstructure formation in a model slurry by the batch sedimentation method. Both experiments and modeling studies were carried out. The results were presented in a recently completed Ph.D. thesis. The third task involved the use of laser confocal microscopy to study the effectiveness of three slurry rheology modifiers. An effective modifier was identified which resulted in lowering the yield stress of the waste simulant. Therefore, the results of this research have led to the basic understanding of the foaming/antifoaming mechanism in waste slurries as well as identification of a rheology modifier, which enhances the processing throughput, and accelerates the DOE mission. The objectives of this research effort were to develop a fundamental understanding of the physico-chemical mechanisms that produced foaming and air entrainment in the DOE High Level (HLW) and Low Activity (LAW) radioactive waste separation and immobilization processes, and to develop and test advanced antifoam/defoaming/rheology modifier agents. Antifoams/rheology modifiers developed from this research ere tested using non-radioactive simulants of the radioactive wastes obtained from Hanford and the Savannah River Site (SRS).

  5. Progress of the High Level Waste Program at the Defense Waste Processing Facility - 13178

    SciTech Connect (OSTI)

    Bricker, Jonathan M.; Fellinger, Terri L.; Staub, Aaron V.; Ray, Jeff W.; Iaukea, John F. [Savannah River Remediation, Aiken, South Carolina, 29808 (United States)] [Savannah River Remediation, Aiken, South Carolina, 29808 (United States)

    2013-07-01T23:59:59.000Z

    The Defense Waste Processing Facility at the Savannah River Site treats and immobilizes High Level Waste into a durable borosilicate glass for safe, permanent storage. The High Level Waste program significantly reduces environmental risks associated with the storage of radioactive waste from legacy efforts to separate fissionable nuclear material from irradiated targets and fuels. In an effort to support the disposition of radioactive waste and accelerate tank closure at the Savannah River Site, the Defense Waste Processing Facility recently implemented facility and flowsheet modifications to improve production by 25%. These improvements, while low in cost, translated to record facility production in fiscal years 2011 and 2012. In addition, significant progress has been accomplished on longer term projects aimed at simplifying and expanding the flexibility of the existing flowsheet in order to accommodate future processing needs and goals. (authors)

  6. Chapter 38 Hazardous Waste Permitting Process (Kentucky)

    Broader source: Energy.gov [DOE]

    This administrative regulation establishes the general provisions for storage, treatment, recycling, or disposal of hazardous waste. It provides information about permits and specific requirements...

  7. Unit operations used to treat process and/or waste streams at nuclear power plants. [R

    SciTech Connect (OSTI)

    Godbee, H.W.; Kibbey, A.H.

    1980-01-01T23:59:59.000Z

    Estimates are given of the annual amounts of each generic type of LLW (i.e., Government and commerical (fuel cycle and non-fuel cycle)) that is generated at LWR plants. Many different chemical engineering unit operations used to treat process and/or waste streams at LWR plants include adsorption, evaporation, calcination, centrifugation, compaction, crystallization, drying, filtration, incineration, reverse osmosis, and solidification of waste residues. The treatment of these various streams and the secondary wet solid wastes thus generated is described. The various treatment options for concentrates or solid wet wastes, and for dry wastes are discussed. Among the dry waste treatment methods are compaction, baling, and incineration, as well as chopping, cutting and shredding. Organic materials (liquids (e.g., oils or solvents) and/or solids), could be incinerated in most cases. The filter sludges, spent resins, and concentrated liquids (e.g., evaporator concentrates) are usually solidified in cement, or urea-formaldehyde or unsaturated polyester resins prior to burial. Incinerator ashes can also be incorporated in these binding agents. Asphalt has not yet been used. This paper presents a brief survey of operational experience at LWRs with various unit operations, including a short discussion of problems and some observations on recent trends.

  8. Proposed Changes to EPA's Transuranic Waste Characterization Approval Process

    SciTech Connect (OSTI)

    Joglekar. R. D.; Feltcorn, E. M.; Ortiz, A. M.

    2003-02-25T23:59:59.000Z

    This paper describes the changes to the waste characterization (WC) approval process proposed in August 2002 by the U.S. Environmental Protection Agency (EPA or the Agency or we). EPA regulates the disposal of transuranic (TRU) waste at the Waste Isolation Pilot Plant (WIPP) repository in Carlsbad, New Mexico. EPA regulations require that waste generator/storage sites seek EPA approval of WC processes used to characterize TRU waste destined for disposal at WIPP. The regulations also require that EPA verify, through site inspections, characterization of each waste stream or group of waste streams proposed for disposal at the WIPP. As part of verification, the Agency inspects equipment, procedures, and interviews personnel to determine if the processes used by a site can adequately characterize the waste in order to meet the waste acceptance criteria for WIPP. The paper discusses EPA's mandate, current regulations, inspection experience, and proposed changes. We expect that th e proposed changes will provide equivalent or improved oversight. Also, they would give EPA greater flexibility in scheduling and conducting inspections, and should clarify the regulatory process of inspections for both Department of Energy (DOE) and the public.

  9. aqueous liquid waste: Topics by E-print Network

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

    Majda 2005-01-01 18 Momentum, Heat, and Neutral Mass Transport in Convective Atmospheric Pressure Plasma-Liquid Systems and Implications for Aqueous Targets CERN Preprints...

  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)

    1989-01-01T23:59:59.000Z

    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. Double Shell Tank (DST) Process Waste Sampling Subsystem Definition Report

    SciTech Connect (OSTI)

    RASMUSSEN, J.H.

    2000-04-25T23:59:59.000Z

    This report defines the Double-Shell Tank (DST) Process Waste Sampling Subsystem (PWSS). This subsystem definition report fully describes and identifies the system boundaries of the PWSS. This definition provides a basis for developing functional, performance, and test requirements (i.e., subsystem specification), as necessary, for the PWSS. The resultant PWSS specification will include the sampling requirements to support the transfer of waste from the DSTs to the Privatization Contractor during Phase 1 of Waste Feed Delivery.

  12. Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

    SciTech Connect (OSTI)

    Digby D. Macdonald; Brian M. Marx; Sejin Ahn; Julio de Ruiz; Balaji Soundararaja; Morgan Smith; and Wendy Coulson

    2008-01-15T23:59:59.000Z

    Various forms of general and localized corrosion represent principal threats to the integrity of DOE liquid waste storage tanks. These tanks, which are of a single wall or double wall design, depending upon their age, are fabricated from welded carbon steel and contain a complex waste-form comprised of NaOH and NaNO{sub 3}, along with trace amounts of phosphate, sulfate, carbonate, and chloride. Because waste leakage can have a profound environmental impact, considerable interest exists in predicting the accumulation of corrosion damage, so as to more effectively schedule maintenance and repair. The different tasks that are being carried out under the current program are as follows: (1) Theoretical and experimental assessment of general corrosion of iron/steel in borate buffer solutions by using electrochemical impedance spectroscopy (EIS), ellipsometry and XPS techniques; (2) Development of a damage function analysis (DFA) which would help in predicting the accumulation of damage due to pitting corrosion in an environment prototypical of DOE liquid waste systems; (3) Experimental measurement of crack growth rate, acoustic emission signals and coupling currents for fracture in carbon and low alloy steels as functions of mechanical (stress intensity), chemical (conductivity), electrochemical (corrosion potential, ECP), and microstructural (grain size, precipitate size, etc) variables in a systematic manner, with particular attention being focused on the structure of the noise in the current and its correlation with the acoustic emissions; (4) Development of fracture mechanisms for carbon and low alloy steels that are consistent with the crack growth rate, coupling current data and acoustic emissions; (5) Inserting advanced crack growth rate models for SCC into existing deterministic codes for predicting the evolution of corrosion damage in DOE liquid waste storage tanks; (6) Computer simulation of the anodic and cathodic activity on the surface of the steel samples in order to exactly predict the corrosion mechanisms; (7) Wavelet analysis of EC noise data from steel samples undergoing corrosion in an environment similar to that of the high level waste storage containers, to extract data pertaining to general, pitting and stress corrosion processes, from the overall data. The Point Defect Model (PDM) is directly applied as the theoretical assessment method for describing the passive film formed on iron/steels. The PDM is used to describe general corrosion in the passive region of iron. In addition, previous work suggests that pit formation is due to the coalescence of cation vacancies at the metal/film interface which would make it possible to use the PDM parameters to predict the onset of pitting. This previous work suggests that once the critical vacancy density is reached, the film ruptures to form a pit. Based upon the kinetic parameters derived for the general corrosion case, two parameters relating to the cation vacancy formation and annihilation can be calculated. These two parameters can then be applied to predict the transition from general to pitting corrosion for iron/mild steels. If cation vacancy coalescence is shown to lead to pitting, it can have a profound effect on the direction of future studies involving the onset of pitting corrosion. The work has yielded a number of important findings, including an unequivocal demonstration of the role of chloride ion in passivity breakdown on nickel in terms of cation vacancy generation within the passive film, the first detection and characterization of individual micro fracture events in stress corrosion cracking, and the determination of kinetic parameters for the generation and annihilation of point defects in the passive film on iron. The existence of coupling between the internal crack environment and the external cathodic environment, as predicted by the coupled environment fracture model (CEFM), has also been indisputably established for the AISI 4340/NaOH system. It is evident from the studies that analysis of coupling current noise is a very sensitive tool f

  13. Process for stabilization of coal liquid fractions

    DOE Patents [OSTI]

    Davies, Geoffrey (Boston, MA); El-Toukhy, Ahmed (Alexandria, EG)

    1987-01-01T23:59:59.000Z

    Coal liquid fractions to be used as fuels are stabilized against gum formation and viscosity increases during storage, permitting the fuel to be burned as is, without further expensive treatments to remove gums or gum-forming materials. Stabilization is accomplished by addition of cyclohexanol or other simple inexpensive secondary and tertiary alcohols, secondary and tertiary amines, and ketones to such coal liquids at levels of 5-25% by weight with respect to the coal liquid being treated. Cyclohexanol is a particularly effective and cost-efficient stabilizer. Other stabilizers are isopropanol, diphenylmethanol, tertiary butanol, dipropylamine, triethylamine, diphenylamine, ethylmethylketone, cyclohexanone, methylphenylketone, and benzophenone. Experimental data indicate that stabilization is achieved by breaking hydrogen bonds between phenols in the coal liquid, thereby preventing or retarding oxidative coupling. In addition, it has been found that coal liquid fractions stabilized according to the invention can be mixed with petroleum-derived liquid fuels to produce mixtures in which gum deposition is prevented or reduced relative to similar mixtures not containing stabilizer.

  14. Method for the simultaneous recovery of radionuclides from liquid radioactive wastes using a solvent

    DOE Patents [OSTI]

    Romanovskiy, Valeriy Nicholiavich (St. Petersburg, RU); Smirnov, Igor V. (St. Petersburg, RU); Babain, Vasiliy A. (St. Petersburg, RU); Todd, Terry A. (Aberdeen, ID); Brewer, Ken N. (Arco, ID)

    2001-01-01T23:59:59.000Z

    The present invention relates to solvents, and methods, for selectively extracting and recovering radionuclides, especially cesium and strontium, rare earths and actinides from liquid radioactive wastes. More specifically, the invention relates to extracting agent solvent compositions comprising complex organoboron compounds, substituted polyethylene glycols, and neutral organophosphorus compounds in a diluent. The preferred solvent comprises a chlorinated cobalt dicarbollide, diphenyl-dibutylmethylenecarbamoylphosphine oxide, PEG-400, and a diluent of phenylpolyfluoroalkyl sulfone. The invention also provides a method of using the invention extracting agents to recover cesium, strontium, rare earths and actinides from liquid radioactive waste.

  15. Solvent for the simultaneous recovery of radionuclides from liquid radioactive wastes

    DOE Patents [OSTI]

    Romanovskiy, Valeriy Nicholiavich (St. Petersburg, RU); Smirnov, Igor V. (St. Petersburg, RU); Babain, Vasiliy A. (St. Petersburg, RU); Todd, Terry A. (Aberdeen, ID); Brewer, Ken N. (Arco, ID)

    2002-01-01T23:59:59.000Z

    The present invention relates to solvents, and methods, for selectively extracting and recovering radionuclides, especially cesium and strontium, rare earths and actinides from liquid radioactive wastes. More specifically, the invention relates to extracting agent solvent compositions comprising complex organoboron compounds, substituted polyethylene glycols, and neutral organophosphorus compounds in a diluent. The preferred solvent comprises a chlorinated cobalt dicarbollide, diphenyl-dibutylmethylenecarbamoylphosphine oxide, PEG-400, and a diluent of phenylpolyfluoroalkyl sulfone. The invention also provides a method of using the invention extracting agents to recover cesium, strontium, rare earths and actinides from liquid radioactive waste.

  16. Description of processes for the immobilization of selected transuranic wastes

    SciTech Connect (OSTI)

    Timmerman, C.L.

    1980-12-01T23:59:59.000Z

    Processed sludge and incinerator-ash wastes contaminated with transuranic (TRU) elements may require immobilization to prevent the release of these elements to the environment. As part of the TRU Waste Immobilization Program sponsored by the Department of Energy (DOE), the Pacific Northwest Laboratory is developing applicable waste-form and processing technology that may meet this need. This report defines and describes processes that are capable of immobilizing a selected TRU waste-stream consisting of a blend of three parts process sludge and one part incinerator ash. These selected waste streams are based on the compositions and generation rates of the waste processing and incineration facility at the Rocky Flats Plant. The specific waste forms that could be produced by the described processes include: in-can melted borosilicate-glass monolith; joule-heated melter borosilicate-glass monolith or marble; joule-heated melter aluminosilicate-glass monolith or marble; joule-heated melter basaltic-glass monolith or marble; joule-heated melter glass-ceramic monolith; cast-cement monolith; pressed-cement pellet; and cold-pressed sintered-ceramic pellet.

  17. Hanford's Simulated Low Activity Waste Cast Stone Processing

    SciTech Connect (OSTI)

    Kim, Young

    2013-08-20T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1996-05-01T23:59:59.000Z

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

  19. Waste-Lithium-Liquid (WLL) Flow Battery for Stationary Energy Storage Applications Youngsik Kim* and Nina MahootcheianAsl

    E-Print Network [OSTI]

    Zhou, Yaoqi

    Waste-Lithium-Liquid (WLL) Flow Battery for Stationary Energy Storage Applications Youngsik Kim in a Waste-Lithium-Liquid (WLL) flow battery that can be used in a stationary energy storage application. Li* and Nina MahootcheianAsl Richard Lugar Center for Renewable Energy, Department of Mechanical Engineering

  20. Retrieval process development and enhancements waste simulant compositions and defensibility

    SciTech Connect (OSTI)

    Powell, M.R.; Golcar, G.R.; Geeting, J.G.H.

    1997-09-01T23:59:59.000Z

    The purpose of this report is to document the physical waste simulant development efforts of the EM-50 Tanks Focus Area at the Hanford Site. Waste simulants are used in the testing and development of waste treatment and handling processes because performing such tests using actual tank waste is hazardous and prohibitively expensive. This document addresses the simulant development work that supports the testing of waste retrieval processes using simulants that mimic certain key physical properties of the tank waste. Development and testing of chemical simulants are described elsewhere. This work was funded through the EM-50 Tanks Focus Area as part of the Retrieval Process Development and Enhancements (RPD&E) Project at the Pacific Northwest National Laboratory (PNNL). The mission of RPD&E is to understand retrieval processes, including emerging and existing processes, gather performance data on those processes, and relate the data to specific tank problems to provide end users with the requisite technical bases to make retrieval and closure decisions. Physical simulants are prepared using relatively nonhazardous and inexpensive materials rather than the chemicals known to be in tank waste. Consequently, only some of the waste properties are matched by the simulant. Deciding which properties need to be matched and which do not requires a detailed knowledge of the physics of the process to be tested using the simulant. Developing this knowledge requires reviews of available literature, consultation with experts, and parametric tests. Once the relevant properties are identified, waste characterization data are reviewed to establish the target ranges for each property. Simulants are then developed that possess the desired ranges of properties.

  1. Technical resource document for assured thermal processing of wastes

    SciTech Connect (OSTI)

    Farrow, R.L.; Fisk, G.A.; Hartwig, C.M.; Hurt, R.H.; Ringland, J.T.; Swansiger, W.A.

    1994-06-01T23:59:59.000Z

    This document is a concise compendium of resource material covering assured thermal processing of wastes (ATPW), an area in which Sandia aims to develop a large program. The ATPW program at Sandia is examining a wide variety of waste streams and thermal processes. Waste streams under consideration include municipal, chemical, medical, and mixed wastes. Thermal processes under consideration range from various incineration technologies to non-incineration processes such as supercritical water oxidation or molten metal technologies. Each of the chapters describes the element covered, discusses issues associated with its further development and/or utilization, presents Sandia capabilities that address these issues, and indicates important connections to other ATPW elements. The division of the field into elements was driven by the team`s desire to emphasize areas where Sandia`s capabilities can lead to major advances and is therefore somewhat unconventional. The report will be valuable to Sandians involved in further ATPW program development.

  2. 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.

    1982-09-01T23:59:59.000Z

    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.

  3. High-Level Waste System Process Interface Description

    SciTech Connect (OSTI)

    d'Entremont, P.D.

    1999-01-14T23:59:59.000Z

    The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment.

  4. Waste heat driven absorption refrigeration process and system

    DOE Patents [OSTI]

    Wilkinson, William H. (Columbus, OH)

    1982-01-01T23:59:59.000Z

    Absorption cycle refrigeration processes and systems are provided which are driven by the sensible waste heat available from industrial processes and other sources. Systems are disclosed which provide a chilled water output which can be used for comfort conditioning or the like which utilize heat from sensible waste heat sources at temperatures of less than 170.degree. F. Countercurrent flow equipment is also provided to increase the efficiency of the systems and increase the utilization of available heat.

  5. Accelerator Production of Tritium project process waste assessment

    SciTech Connect (OSTI)

    Carson, S.D.; Peterson, P.K.

    1995-09-01T23:59:59.000Z

    DOE has made a commitment to compliance with all applicable environmental regulatory requirements. In this respect, it is important to consider and design all tritium supply alternatives so that they can comply with these requirements. The management of waste is an integral part of this activity and it is therefore necessary to estimate the quantities and specific wastes that will be generated by all tritium supply alternatives. A thorough assessment of waste streams includes waste characterization, quantification, and the identification of treatment and disposal options. The waste assessment for APT has been covered in two reports. The first report was a process waste assessment (PWA) that identified and quantified waste streams associated with both target designs and fulfilled the requirements of APT Work Breakdown Structure (WBS) Item 5.5.2.1. This second report is an expanded version of the first that includes all of the data of the first report, plus an assessment of treatment and disposal options for each waste stream identified in the initial report. The latter information was initially planned to be issued as a separate Waste Treatment and Disposal Options Assessment Report (WBS Item 5.5.2.2).

  6. PLUTONIUM FINISHING PLANT (PFP) 241-Z LIQUID WASTE TREATMENT FACILITY DEACTIVATION AND DEMOLITION

    SciTech Connect (OSTI)

    JOHNSTON GA

    2008-01-15T23:59:59.000Z

    Fluor Hanford, Inc. (FH) is proud to submit the Plutonium Finishing Plant (PFP) 241-Z liquid Waste Treatment Facility Deactivation and Demolition (D&D) Project for consideration by the Project Management Institute as Project of the Year for 2008. The decommissioning of the 241-Z Facility presented numerous challenges, many of which were unique with in the Department of Energy (DOE) Complex. The majority of the project budget and schedule was allocated for cleaning out five below-grade tank vaults. These highly contaminated, confined spaces also presented significant industrial safety hazards that presented some of the most hazardous work environments on the Hanford Site. The 241-Z D&D Project encompassed diverse tasks: cleaning out and stabilizing five below-grade tank vaults (also called cells), manually size-reducing and removing over three tons of process piping from the vaults, permanently isolating service utilities, removing a large contaminated chemical supply tank, stabilizing and removing plutonium-contaminated ventilation ducts, demolishing three structures to grade, and installing an environmental barrier on the demolition site . All of this work was performed safely, on schedule, and under budget. During the deactivation phase of the project between November 2005 and February 2007, workers entered the highly contaminated confined-space tank vaults 428 times. Each entry (or 'dive') involved an average of three workers, thus equaling approximately 1,300 individual confined -space entries. Over the course of the entire deactivation and demolition period, there were no recordable injuries and only one minor reportable skin contamination. The 241-Z D&D Project was decommissioned under the provisions of the 'Hanford Federal Facility Agreement and Consent Order' (the Tri-Party Agreement or TPA), the 'Resource Conservation and Recovery Act of 1976' (RCRA), and the 'Comprehensive Environmental Response, Compensation, and Liability Act of 1980' (CERCLA). The project completed TPA Milestone M-083-032 to 'Complete those activities required by the 241-Z Treatment and Storage Unit's RCRA Closure Plan' four years and seven months ahead of this legally enforceable milestone. In addition, the project completed TPA Milestone M-083-042 to 'Complete transition and dismantlement of the 241-2 Waste Treatment Facility' four years and four months ahead of schedule. The project used an innovative approach in developing the project-specific RCRA closure plan to assure clear integration between the 241-Z RCRA closure activities and ongoing and future CERCLA actions at PFP. This approach provided a regulatory mechanism within the RCRA closure plan to place segments of the closure that were not practical to address at this time into future actions under CERCLA. Lessons learned from th is approach can be applied to other closure projects within the DOE Complex to control scope creep and mitigate risk. A paper on this topic, entitled 'Integration of the 241-Z Building D and D Under CERCLA with RCRA Closure at the PFP', was presented at the 2007 Waste Management Conference in Tucson, Arizona. In addition, techniques developed by the 241-Z D&D Project to control airborne contamination, clean the interior of the waste tanks, don and doff protective equipment, size-reduce plutonium-contaminated process piping, and mitigate thermal stress for the workers can be applied to other cleanup activities. The project-management team developed a strategy utilizing early characterization, targeted cleanup, and close coordination with PFP Criticality Engineering to significantly streamline the waste- handling costs associated with the project . The project schedule was structured to support an early transition to a criticality 'incredible' status for the 241-Z Facility. The cleanup work was sequenced and coordinated with project-specific criticality analysis to allow the fissile material waste being generated to be managed in a bulk fashion, instead of individual waste packages. This approach negated the need for real-time assay of individ

  7. Zone Freezing Study for Pyrochemical Process Waste Minimization

    SciTech Connect (OSTI)

    Ammon Williams

    2012-05-01T23:59:59.000Z

    Pyroprocessing technology is a non-aqueous separation process for treatment of used nuclear fuel. At the heart of pyroprocessing lies the electrorefiner, which electrochemically dissolves uranium from the used fuel at the anode and deposits it onto a cathode. During this operation, sodium, transuranics, and fission product chlorides accumulate in the electrolyte salt (LiCl-KCl). These contaminates change the characteristics of the salt overtime and as a result, large volumes of contaminated salt are being removed, reprocessed and stored as radioactive waste. To reduce the storage volumes and improve recycling process for cost minimization, a salt purification method called zone freezing has been proposed at Korea Atomic Energy Research Institute (KAERI). Zone freezing is melt crystallization process similar to the vertical Bridgeman method. In this process, the eutectic salt is slowly cooled axially from top to bottom. As solidification occurs, the fission products are rejected from the solid interface and forced into the liquid phase. The resulting product is a grown crystal with the bulk of the fission products near the bottom of the salt ingot, where they can be easily be sectioned and removed. Despite successful feasibility report from KAERI on this process, there were many unexplored parameters to help understanding and improving its operational routines. Thus, this becomes the main motivation of this proposed study. The majority of this work has been focused on the CsCl-LiCl-KCl ternary salt. CeCl3-LiCl-KCl was also investigated to check whether or not this process is feasible for the trivalent species—surrogate for rare-earths and transuranics. For the main part of the work, several parameters were varied, they are: (1) the retort advancement rate—1.8, 3.2, and 5.0 mm/hr, (2) the crucible lid configurations—lid versus no-lid, (3) the amount or size of mixture—50 and 400 g, (4) the composition of CsCl in the salt—1, 3, and 5 wt%, and (5) the temperature differences between the high and low furnace zones—200 and 300 ?C. During each experiment, the temperatures at selected locations around the crucible were measured and recorded to provide temperature profiles. Following each experiment, samples were collected and elemental analysis was done to determine the composition of iii the salt. Several models—non-mixed, well-mixed, Favier, and hybrid—were explored to describe the zone freezing process. For CsCl-LiCl-KCl system, experimental results indicate that through this process up to 90% of the used salt can be recycled, effectively reducing waste volume by a factor of ten. The optimal configuration was found to be a 5.0 mm/hr rate with a lid configuration and a ?T of 200°C. The larger 400 g mixtures had recycle percentages similar to the 50 g mixtures; however, the throughput per time was greater for the 400 g case. As a result, the 400 g case is recommended. For the CeCl3-LiCl-KCl system, the result implies that it is possible to use this process to separate the rare-earth and transuranics chlorides. Different models were applied to only CsCl ternary system. The best fit model was the hybrid model as a result of a solute transport transition from non- mixed to well-mixed throughout the growing process.

  8. Electromagnetic mixed waste processing system for asbestos decontamination

    SciTech Connect (OSTI)

    Kasevich, R.S.; Vaux, W. [KAI Technologies, Inc., Portsmouth, NH (United States); Ulerich, N. [Westinghouse Electric Corp., Pittsburgh, PA (United States). Science and Technology Center; Nocito, T. [Ohio DSI Corporation, New York (New York)

    1996-12-31T23:59:59.000Z

    The overall objective of this three-phase program is to develop an integrated process for treating asbestos-containing material that is contaminated with radioactive and hazardous constituents. The integrated process will attempt to minimize processing and disposal costs. The objectives of Phase 1 were to establish the technical feasibility of asbestos decomposition, inorganic radionuclide nd heavy metal removal, and organic volatilization. Phase 1 resulted in the successful bench-scale demonstration of the elements required to develop a mixed waste treatment process for asbestos-containing material (ACM) contaminated with radioactive metals, heavy metals, and organics. Using the Phase 1 data, a conceptual process was developed. The Phase 2 program, currently in progress, is developing an integrated system design for ACM waste processing. The Phase 3 program will target demonstration of the mixed waste processing system at a DOE facility. The electromagnetic mixed waste processing system employs patented technologies to convert DOE asbestos to a non-hazardous, radionuclide-free, stable waste. The dry, contaminated asbestos is initially heated with radiofrequency energy to remove organic volatiles. Second,the radionuclides are removed by solvent extraction coupled with ion exchange solution treatment. Third, the ABCOV method converts the asbestos to an amorphous silica suspension at low temperature (100{degrees}C). Finally the amorphous silica is solidified for disposal.

  9. Groundwater impact assessment report for the 1325-N Liquid Waste Disposal Facility

    SciTech Connect (OSTI)

    Alexander, D.J.; Johnson, V.G.

    1993-09-01T23:59:59.000Z

    In 1943 the Hanford Site was chosen as a location for the Manhattan Project to produce plutonium for use in nuclear weapons. The 100-N Area at Hanford was used from 1963 to 1987 for a dual-purpose, plutonium production and steam generation reactor and related operational support facilities (Diediker and Hall 1987). In November 1989, the reactor was put into dry layup status. During operations, chemical and radioactive wastes were released into the area soil, air, and groundwater. The 1325-N LWDF was constructed in 1983 to replace the 1301-N Liquid Waste Disposal Facility (1301-N LWDF). The two facilities operated simultaneously from 1983 to 1985. The 1301-N LWDF was retired from use in 1985 and the 1325-N LWDF continued operation until April 1991, when active discharges to the facility ceased. Effluent discharge to the piping system has been controlled by administrative means. This report discusses ground water contamination resulting from the 1325-N Liquid Waste Disposal facility.

  10. Waste Treatment Technology Process Development Plan For Hanford Waste Treatment Plant Low Activity Waste Recycle

    SciTech Connect (OSTI)

    McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.

    2013-08-29T23:59:59.000Z

    The purpose of this Process Development Plan is to summarize the objectives and plans for the technology development activities for an alternative path for disposition of the recycle stream that will be generated in the Hanford Waste Treatment Plant Low Activity Waste (LAW) vitrification facility (LAW Recycle). This plan covers the first phase of the development activities. The baseline plan for disposition of this stream is to recycle it to the WTP Pretreatment Facility, where it will be concentrated by evaporation and returned to the LAW vitrification facility. Because this stream contains components that are volatile at melter temperatures and are also problematic for the glass waste form, they accumulate in the Recycle stream, exacerbating their impact on the number of LAW glass containers. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and reducing the halides in the Recycle is a key component of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, this stream does not have a proven disposition path, and resolving this gap becomes vitally important. This task seeks to examine the impact of potential future disposition of this stream in the Hanford tank farms, and to develop a process that will remove radionuclides from this stream and allow its diversion to another disposition path, greatly decreasing the LAW vitrification mission duration and quantity of glass waste. The origin of this LAW Recycle stream will be from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover or precipitates of scrubbed components (e.g. carbonates). The soluble components are mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and will not be available until the WTP begins operation, causing uncertainty in its composition, particularly the radionuclide content. This plan will provide an estimate of the likely composition and the basis for it, assess likely treatment technologies, identify potential disposition paths, establish target treatment limits, and recommend the testing needed to show feasibility. Two primary disposition options are proposed for investigation, one is concentration for storage in the tank farms, and the other is treatment prior to disposition in the Effluent Treatment Facility. One of the radionuclides that is volatile and expected to be in high concentration in this LAW Recycle stream is Technetium-99 ({sup 99}Tc), a long-lived radionuclide with a half-life of 210,000 years. Technetium will not be removed from the aqueous waste in the Hanford Waste Treatment and Immobilization Plant (WTP), and will primarily end up immobilized in the LAW glass, which will be disposed in the Integrated Disposal Facility (IDF). Because {sup 99}Tc has a very long half-life and is highly mobile, it is the largest dose contributor to the Performance Assessment (PA) of the IDF. Other radionuclides that are also expected to be in appreciable concentration in the LAW Recycle are {sup 129}I, {sup 90}Sr, {sup 137}Cs, and {sup 241}Am. The concentrations of these radionuclides in this stream will be much lower than in the LAW, but they will still be higher than limits for some of the other disposition pathways currently available. Although the baseline process will recycle this stream to the Pretreatment Facility, if the LAW facility begins operation first, this stream will not have a disposition path internal to WTP. One potential solution is to return the stream to the tank farms where it can be evaporated in the 242-A evaporator, or perhaps deploy an auxiliary evaporator to concentrate it prior to return to the tank farms. In either case, testing is needed to evaluat

  11. Ionic Liquids for Utilization of Waste Heat from Distributed Power Generation Systems

    SciTech Connect (OSTI)

    Joan F. Brennecke; Mihir Sen; Edward J. Maginn; Samuel Paolucci; Mark A. Stadtherr; Peter T. Disser; Mike Zdyb

    2009-01-11T23:59:59.000Z

    The objective of this research project was the development of ionic liquids to capture and utilize waste heat from distributed power generation systems. Ionic Liquids (ILs) are organic salts that are liquid at room temperature and they have the potential to make fundamental and far-reaching changes in the way we use energy. In particular, the focus of this project was fundamental research on the potential use of IL/CO2 mixtures in absorption-refrigeration systems. Such systems can provide cooling by utilizing waste heat from various sources, including distributed power generation. The basic objectives of the research were to design and synthesize ILs appropriate for the task, to measure and model thermophysical properties and phase behavior of ILs and IL/CO2 mixtures, and to model the performance of IL/CO2 absorption-refrigeration systems.

  12. IMPACT OF THE SMALL COLUMN ION EXCHANGE PROCESS ON THE DEFENSE WASTE PROCESSING FACILITY - 12112

    SciTech Connect (OSTI)

    Koopman, D.; Lambert, D.; Fox, K.; Stone, M.

    2011-11-07T23:59:59.000Z

    The Savannah River Site (SRS) is investigating the deployment of a parallel technology to the Salt Waste Processing Facility (SWPF, presently under construction) to accelerate high activity salt waste processing. The proposed technology combines large waste tank strikes of monosodium titanate (MST) to sorb strontium and actinides with two ion exchange columns packed with crystalline silicotitanate (CST) resin to sorb cesium. The new process was designated Small Column Ion Exchange (SCIX), since the ion exchange columns were sized to fit within a waste storage tank riser. Loaded resins are to be combined with high activity sludge waste and fed to the Defense Waste Processing Facility (DWPF) for incorporation into the current glass waste form. Decontaminated salt solution produced by SCIX will be fed to the SRS Saltstone Facility for on-site immobilization as a grout waste form. Determining the potential impact of SCIX resins on DWPF processing was the basis for this study. Accelerated salt waste treatment is projected to produce a significant savings in the overall life cycle cost of waste treatment at SRS.

  13. Sorbent Testing For Solidification of Process Waste streams from the Radiochemical Engineering Development Center at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Bickford, J. [MSE Technology Applications, Inc., MT (United States); Taylor, P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    2007-07-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) tasked MSE Technology Applications, Inc. (MSE) to evaluate sorbents identified by Oak Ridge National Laboratory (ORNL) to solidify the radioactive liquid organic waste from the Radiochemical Engineering Development Center (REDC) at ORNL. REDC recovers and purifies heavy elements (berkelium, californium, einsteinium, and fermium) from irradiated targets for research and industrial applications. Both organic and aqueous waste streams are discharged from REDC. The organic waste is generated from the plutonium/uranium extraction (Purex), Cleanex, and Pubex processes. 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. MSE had also been tasked to test a grouting formula for the aqueous waste stream that includes radioactive shielding material. The aqueous waste is a mixture of the raffinate streams from the various extraction processes plus the caustic solution that is used to dissolve the aluminum cladding from the irradiated targets. (authors)

  14. A delayed-neutron monitor for a liquid-waste stream with high gamma-ray intensity

    SciTech Connect (OSTI)

    Rinard, P.M.; Crane, T.W.; Van Lyssel, T.; Kroncke, K.M.; Schneider, C.M.; Bourret, S.C. (Los Alamos National Lab., NM (USA))

    1989-01-01T23:59:59.000Z

    An instrument has been built to monitor the uranium concentration in a liquid-waste stream to avoid a criticality accident in a downstream holding tank. The measurement technique is based on the production and counting of delayed neutrons using the shuffler'' process because the waste contains enough fission products to produce a gamma-ray dose rate of 10 R/h on the surface of the assay tank. The design goal was a sensitivity of 0.034 g/L (1{sigma} = 10%) in 100 s as the stream flows at 80 L/h through the assay chamber. The instrument is to run unattended for at least three months; during this time it it to transmit assay results to the plant computer and generate warnings and alarms when necessary.

  15. THE ROLE OF LIQUID WASTE PRETREATMENT TECHNOLOGIES IN SOLVING THE DOE CLEAN-UP MISSION

    SciTech Connect (OSTI)

    Wilmarth, B; Sheryl Bush, S

    2008-10-31T23:59:59.000Z

    The objective of this report is to describe the pretreatment solutions that allow treatment to be tailored to specific wastes, processing ahead of the completion schedules for the main treatment facilities, and reduction of technical risks associated with future processing schedules. Wastes stored at Hanford and Savannah River offer challenging scientific and engineering tasks. At both sites, space limitations confound the ability to effectively retrieve and treat the wastes. Additionally, the radiation dose to the worker operating and maintaining the radiochemical plants has a large role in establishing the desired radioactivity removal. However, the regulatory requirements to treat supernatant and saltcake tank wastes differ at the two sites. Hanford must treat and remove radioactivity from the tanks based on the TriParty Agreement and Waste Incidental to Reprocessing (WIR) documentation. These authorizing documents do not specify treatment technologies; rather, they specify endstate conditions. Dissimilarly, Waste Determinations prepared at SRS in accordance with Section 3116 of the 2005 National Defense Authorization Act along with state operating permits establish the methodology and amounts of radioactivity that must be removed and may be disposed of in South Carolina. After removal of entrained solids and site-specific radionuclides, supernatant and saltcake wastes are considered to be low activity waste (LAW) and are immobilized in glass and disposed of at the Hanford Site Integrated Disposal Facility (IDF) or formulated into a grout for disposal at the Savannah River Site Saltstone Disposal Facility. Wastes stored at the Hanford Site or SRS comprise saltcake, supernate, and sludges. The supernatant and saltcake waste fractions contain primarily sodium salts, metals (e.g., Al, Cr), cesium-137 (Cs-137), technetium-99 (Tc-99) and entrained solids containing radionuclides such as strontium-90 (Sr-90) and transuranic elements. The sludges contain many of the transition metal hydroxides that precipitate when the spent acidic process solutions are rendered alkaline with sodium hydroxide. The sludges contain Sr-90 and transuranic elements. The wastes stored at each site have been generated and stored for over fifty years. Although the majority of the wastes were generated to support nuclear weapons production and reprocessing, the wastes differ substantially between the sites. Table 5 shows the volumes and total radioactivity (including decay daughters) of the waste phases stored in tanks at each site. At Hanford, there are 177 tanks that contain 56.5 Mgal of waste. SRS has 51 larger tanks, of which 2 are closed, that contain 36.5 Mgal. Mainly due to recovery operations, the waste stored at Hanford has less total curies than that stored at Savannah River. The total radioactivity of the Hanford wastes contains approximately 190 MCi, and the total radioactivity of the Savannah River wastes contains 400 MCi.

  16. RECENT PROCESS AND EQUIPMENT IMPROVEMENTS TO INCREASE HIGH LEVEL WASTE THROUGHPUT AT THE DEFENSE WASTE PROCESSING FACILITY

    SciTech Connect (OSTI)

    Odriscoll, R; Allan Barnes, A; Jim Coleman, J; Timothy Glover, T; Robert Hopkins, R; Dan Iverson, D; Jeff Leita, J

    2008-01-15T23:59:59.000Z

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF) began stabilizing high level waste (HLW) in a glass matrix in 1996. Over the past few years, there have been several process and equipment improvements at the DWPF to increase the rate at which the high level waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process to upsets, thereby minimizing downtime and increasing production. Improvements due to optimization of waste throughput with increased HLW loading of the glass resulted in a 6% waste throughput increase based upon operational efficiencies. Improvements in canister production include the pour spout heated bellows liner (5%), glass surge (siphon) protection software (2%), melter feed pump software logic change to prevent spurious interlocks of the feed pump with subsequent dilution of feed stock (2%) and optimization of the steam atomized scrubber (SAS) operation to minimize downtime (3%) for a total increase in canister production of 12%. A number of process recovery efforts have allowed continued operation. These include the off gas system pluggage and restoration, slurry mix evaporator (SME) tank repair and replacement, remote cleaning of melter top head center nozzle, remote melter internal inspection, SAS pump J-Tube recovery, inadvertent pour scenario resolutions, dome heater transformer bus bar cooling water leak repair and new Infra-red camera for determination of glass height in the canister are discussed.

  17. Materials selection for process equipment in the Hanford waste vitrification plant

    SciTech Connect (OSTI)

    Elmore, M R; Jensen, G A

    1991-07-01T23:59:59.000Z

    The Hanford Waste Vitrification Plant (HWVP) is being designed to vitrify defense liquid high-level wastes and transuranic wastes stored at Hanford. The HWVP Functional Design Criteria (FDC) requires that materials used for fabrication of remote process equipment and piping in the facility be compatible with the expected waste stream compositions and process conditions. To satisfy FDC requirements, corrosion-resistant materials have been evaluated under simulated HWVP-specific conditions and recommendations have been made for HWVP applications. The materials recommendations provide to the project architect/engineer the best available corrosion rate information for the materials under the expected HWVP process conditions. Existing data and sound engineering judgement must be used and a solid technical basis must be developed to define an approach to selecting suitable construction materials for the HWVP. This report contains the strategy, approach, criteria, and technical basis developed for selecting materials of construction. Based on materials testing specific to HWVP and on related outside testing, this report recommends for constructing specific process equipment and identifies future testing needs to complete verification of the performance of the selected materials. 30 refs., 7 figs., 11 tabs.

  18. Materials evaluation programs at the Defense Waste Processing Facility

    SciTech Connect (OSTI)

    Gee, J.T.; Iverson, D.C.; Bickford, D.F.

    1992-12-31T23:59:59.000Z

    The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950s to produce nuclear materials in support of the national defense effort. About 83 million gallons of high-level waste produced since operations began has been consolidated by evaporation into 33 million gallons at the waste tank farm. The Department of Energy authorized the construction of the Defense Waste Processing Facility (DWPF), the function of which is to immobilize the waste as a durable borosilicate glass contained in stainless steel canisters prior to the placement of the canisters in a federal repository. The DWPF is now mechanically complete and is undergoing commissioning and run-in activities. A brief description of the DWPF process is provided.

  19. Materials evaluation programs at the Defense Waste Processing Facility

    SciTech Connect (OSTI)

    Gee, J.T.; Iverson, D.C.; Bickford, D.F.

    1992-01-01T23:59:59.000Z

    The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950s to produce nuclear materials in support of the national defense effort. About 83 million gallons of high-level waste produced since operations began has been consolidated by evaporation into 33 million gallons at the waste tank farm. The Department of Energy authorized the construction of the Defense Waste Processing Facility (DWPF), the function of which is to immobilize the waste as a durable borosilicate glass contained in stainless steel canisters prior to the placement of the canisters in a federal repository. The DWPF is now mechanically complete and is undergoing commissioning and run-in activities. A brief description of the DWPF process is provided.

  20. Process to separate transuranic elements from nuclear waste

    DOE Patents [OSTI]

    Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

    1988-07-12T23:59:59.000Z

    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). 2 figs.

  1. Process to separate transuranic elements from nuclear waste

    DOE Patents [OSTI]

    Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

    1989-03-21T23:59:59.000Z

    A process is described 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). 2 figs.

  2. Process for blending coal with water immiscible liquid

    DOE Patents [OSTI]

    Heavin, Leonard J. (Olympia, WA); King, Edward E. (Gig Harbor, WA); Milliron, Dennis L. (Lacey, WA)

    1982-10-26T23:59:59.000Z

    A continuous process for blending coal with a water immiscible liquid produces a uniform, pumpable slurry. Pulverized raw feed coal and preferably a coal derived, water immiscible liquid are continuously fed to a blending zone (12 and 18) in which coal particles and liquid are intimately admixed and advanced in substantially plug flow to form a first slurry. The first slurry is withdrawn from the blending zone (12 and 18) and fed to a mixing zone (24) where it is mixed with a hot slurry to form the pumpable slurry. A portion of the pumpable slurry is continuously recycled to the blending zone (12 and 18) for mixing with the feed coal.

  3. High magnetic field processing of liquid crystalline polymers

    DOE Patents [OSTI]

    Smith, Mark E. (Los Alamos, NM); Benicewicz, Brian C. (Los Alamos, NM); Douglas, Elliot P. (Los Alamos, NM)

    1998-01-01T23:59:59.000Z

    A process of forming bulk articles of oriented liquid crystalline thermoset material, the material characterized as having an enhanced tensile modulus parallel to orientation of an applied magnetic field of at least 25 percent greater than said material processed in the absence of a magnetic field, by curing a liquid crystalline thermoset precursor within a high strength magnetic field of greater than about 2 Tesla, is provided, together with a resultant bulk article of a liquid crystalline thermoset material, said material processed in a high strength magnetic field whereby said material is characterized as having a tensile modulus parallel to orientation of said field of at least 25 percent greater than said material processed in the absence of a magnetic field.

  4. High magnetic field processing of liquid crystalline polymers

    DOE Patents [OSTI]

    Smith, M.E.; Benicewicz, B.C.; Douglas, E.P.

    1998-11-24T23:59:59.000Z

    A process of forming bulk articles of oriented liquid crystalline thermoset material, the material characterized as having an enhanced tensile modulus parallel to orientation of an applied magnetic field of at least 25 percent greater than said material processed in the absence of a magnetic field, by curing a liquid crystalline thermoset precursor within a high strength magnetic field of greater than about 2 Tesla, is provided, together with a resultant bulk article of a liquid crystalline thermoset material, said material processed in a high strength magnetic field whereby said material is characterized as having a tensile modulus parallel to orientation of said field of at least 25 percent greater than said material processed in the absence of a magnetic field.

  5. Crystallization process to reduce NORM-containing waste

    SciTech Connect (OSTI)

    Hayden, C.G. [Inst. of Gas Technology, Chicago, IL (United States); Kraemer, T.F.

    1995-11-01T23:59:59.000Z

    This paper describes a process of Naturally Occurring Radioactive Material (NORM) waste reduction for scales, sludges, and soils. The process involves dissolution and fractional crystallization steps that concentrate the radioactive material into a small mass of barite. The concentration of radium in the product, barite, can be increased or decreased. The NORM-containing barite product is suitable for slurry injection into sandstone formations.

  6. TECHNOLOGY SUMMARY ADVANCING TANK WASTE RETRIEVAL AND PROCESSING

    SciTech Connect (OSTI)

    SAMS TL; MENDOZA RE

    2010-08-11T23:59:59.000Z

    This technology overview provides a high-level summary of technologies being investigated and developed by Washington River Protection Solutions (WRPS) to advance Hanford Site tank waste retrieval and processing. Technology solutions are outlined, along with processes and priorities for selecting and developing them.

  7. TECHNOLOGY SUMMARY ADVANCING TANK WASTE RETREIVAL AND PROCESSING

    SciTech Connect (OSTI)

    SAMS TL

    2010-07-07T23:59:59.000Z

    This technology overview provides a high-level summary of technologies being investigated and developed by Washington River Protection Solutions (WRPS) to advance Hanford Site tank waste retrieval and processing. Technology solutions are outlined, along with processes and priorities for selecting and developing them.

  8. The Polymers for Liquid Radioactive Waste Solidification: a Lost Chapter in the History of Engineering or a Step Forward? - 13529

    SciTech Connect (OSTI)

    Pokhitonov, Yury [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation)] [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation); Kelley, Dennis [Pacific Nuclear Solutions, Indianapolis, Indiana (United States)] [Pacific Nuclear Solutions, Indianapolis, Indiana (United States)

    2013-07-01T23:59:59.000Z

    Ideas on the application of polymers for the liquid radioactive waste immobilization go a way back, and the first studies in the area were published 30-40 years ago. One should admit that regardless of the fairly large number of publications appeared in the past years currently the interest in this work came down greatly. It was the successful assimilation and worldwide implementation of the LRW cementation technology caused a slump in the interest in polymers. But today it's safe to say that the situation slowly changes, particularly due to the market appearance of the high-tech polymers manufactured by Nochar Company, and unique properties of these polymers gradually raise the demand in various countries. The results of multiple experiments performed with the simulated solutions have passed the comprehensive tests with actual waste. The economic effect from the implementation of the new technology is defined by the volume reduction of waste coming onto the repository, by the decline in the cost of transportation and of the repository construction on account of cutting down the construction volume. Interesting results have been obtained during the search for the technical decisions that would allow using the polymer materials in the processing technology of the industrial toxic waste. One more promising area of the possible application of polymers should be pointed out. It is the application of polymer materials as the assets for the emergency damage control when the advantages of the polymers become obvious. (authors)

  9. 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.

    1991-07-01T23:59:59.000Z

    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.

  10. SAVANNAH RIVER SITE INCIPIENT SLUDGE MIXING IN RADIOACTIVE LIQUID WASTE STORAGE TANKS DURING SALT SOLUTION BLENDING

    SciTech Connect (OSTI)

    Leishear, R.; Poirier, M.; Lee, S.; Steeper, T.; Fowley, M.; Parkinson, K.

    2011-01-12T23:59:59.000Z

    This paper is the second in a series of four publications to document ongoing pilot scale testing and computational fluid dynamics (CFD) modeling of mixing processes in 85 foot diameter, 1.3 million gallon, radioactive liquid waste, storage tanks at Savannah River Site (SRS). Homogeneous blending of salt solutions is required in waste tanks. Settled solids (i.e., sludge) are required to remain undisturbed on the bottom of waste tanks during blending. Suspension of sludge during blending may potentially release radiolytically generated hydrogen trapped in the sludge, which is a safety concern. The first paper (Leishear, et. al. [1]) presented pilot scale blending experiments of miscible fluids to provide initial design requirements for a full scale blending pump. Scaling techniques for an 8 foot diameter pilot scale tank were also justified in that work. This second paper describes the overall reasons to perform tests, and documents pilot scale experiments performed to investigate disturbance of sludge, using non-radioactive sludge simulants. A third paper will document pilot scale CFD modeling for comparison to experimental pilot scale test results for both blending tests and sludge disturbance tests. That paper will also describe full scale CFD results. The final paper will document additional blending test results for stratified layers in salt solutions, scale up techniques, final full scale pump design recommendations, and operational recommendations. Specifically, this paper documents a series of pilot scale tests, where sludge simulant disturbance due to a blending pump or transfer pump are investigated. A principle design requirement for a blending pump is UoD, where Uo is the pump discharge nozzle velocity, and D is the nozzle diameter. Pilot scale test results showed that sludge was undisturbed below UoD = 0.47 ft{sup 2}/s, and that below UoD = 0.58 ft{sup 2}/s minimal sludge disturbance was observed. If sludge is minimally disturbed, hydrogen will not be released. Installation requirements were also determined for a transfer pump which will remove tank contents, and which is also required to not disturb sludge. Testing techniques and test results for both types of pumps are presented.

  11. New Standards in Liquid Waste Treatment at Fukushima Dai-ichi - 13134

    SciTech Connect (OSTI)

    Sylvester, Paul; Milner, Tim; Ruffing, Jennifer; Poole, Scott [EnergySolutions, 100 Center Point Circle, Suite 100, Center Point II, Columbia, SC 29210 (United States)] [EnergySolutions, 100 Center Point Circle, Suite 100, Center Point II, Columbia, SC 29210 (United States); Townson, Paul; Jensen, Jesse [EnergySolutions, 2345 Stevens Drive, Suite 240, Richland, WA 99354 (United States)] [EnergySolutions, 2345 Stevens Drive, Suite 240, Richland, WA 99354 (United States)

    2013-07-01T23:59:59.000Z

    The earthquake and tsunami on March 11, 2011 severely damaged the Fukushima Dai-ichi nuclear plant leading to the most severe nuclear incident since Chernobyl. Ongoing operations to cool the damaged reactors at the site have led to the generation of highly radioactive coolant water. This is currently mainly treated to remove Cs-137 and Cs-134 and passed through a reverse osmosis (RO) unit to reduce the salinity before being cycled back to the reactors. Because only the Cs isotopes are removed, the RO reject water still contains many radioactive isotopes and this has led to the accumulation of over 200,000 cubic meters (52 million gallons) of extremely contaminated water which is currently stored on site in tanks. EnergySolutions, in partnership with Toshiba, were contracted to develop a system to reduce 62 isotopes in this waste down to allowable levels. This was a significant technical challenge given the high background salt content of the wastewater, the variation in aqueous chemistry of the radioactive isotopes and the presence of non-active competing ions (e.g. Ca and Mg) which inhibit the removal of isotopes such as Sr-89 and Sr-90. Extensive testing was performed to design a suitable system that could meet the required decontamination goals. These tests were performed over a 6 month period at facilities available in the nearby Fukushima Dai-ni laboratory using actual waste samples. This data was then utilized to design a Multi Radioactive Nuclides Removal System (MRRS) for Fukushima which is a modified version of EnergySolutions' proprietary Advanced Liquid Processing System (ALPS)'. The stored tank waste is fed into a preliminary precipitation system where iron flocculation is performed to remove a number of isotopes, including Sb-125, Ru-106, Mn-54 and Co-60. The supernatant is then fed into a second precipitation tank where the pH is adjusted and the bulk of the Mg, Ca and Sr precipitated out as carbonates and hydroxides. After passing through a cross-flow ultrafiltration membrane, the permeate then goes through a total of 14 fixed ion exchange and adsorbent columns followed by a disposable polishing column to polish the residual isotopes down to allowable levels. At the end of the system, the effluent is filtered for a final time to removal any particulates that may have been picked up from the media columns and then stored prior to analysis. (authors)

  12. Determination of vapor-liquid equilibrium data and decontamination factors needed for the development of evaporator technology for use in volume reduction of radioactive waste streams

    SciTech Connect (OSTI)

    Betts, S.E. [California Univ., Santa Barbara, CA (United States)

    1993-10-01T23:59:59.000Z

    A program is currently in progress at Argonne National Laboratory to evaluate and develop evaporator technology for concentrating radioactive waste streams. By concentrating radioactive waste streams, disposal costs can be significantly reduced. To effectively reduce the volume of waste, the evaporator must achieve high decontamination factors so that the distillate is sufficiently free of radioactive material. One technology that shows a great deal of potential for this application is being developed by LICON, Inc. In this program, Argonne plans to apply LICON`s evaporator designs to the processing of radioactive solutions. Concepts that need to be incorporated into the design of the evaporator include, criticality safety, remote operation and maintenance, and materials of construction. To design an effective process for concentrating waste streams, both solubility and vapor-liquid equilibrium data are needed. The key issue, however, is the high decontamination factors that have been demonstrated by this equipment. Two major contributions were made to this project. First, a literature survey was completed to obtain available solubility and vapor-liquid equilibrium data. Some vapor-liquid data necessary for the project but not available in the literature was obtained experimentally. Second, the decontamination factor for the evaporator was determined using neutron activation analysis (NAA).

  13. Identification and differentiation of individual beta emitters in waste mixtures by liquid scintillation spectrometry

    E-Print Network [OSTI]

    Siskel, Robin Lynn

    1988-01-01T23:59:59.000Z

    carbon-14, tritium, and iodine-125 liquid scintillation wastes, provided that the activity and isotopes present can be documented. This legislation has generated a significant interest in developing a quick, cost efficient method of identificatior... to reference data was the most practical method of unknown identification, and fully complies with regulatory requirements for relatively unquenched sample mixtures of two isotopes. Accurate isotopic identification and differentiation was found...

  14. International Best Practices for Pre-Processing and Co-Processing Municipal Solid Waste and Sewage Sludge in the Cement Industry

    E-Print Network [OSTI]

    Hasanbeigi, Ali

    2013-01-01T23:59:59.000Z

    Co-processing Municipal Solid Waste and Sewage Sludge in theno date. “Integrated Solid Waste Management. ” Presentationincineration of Municipal Solid Waste in Cement Industry. :

  15. Direction of CRT waste glass processing: Electronics recycling industry communication

    SciTech Connect (OSTI)

    Mueller, Julia R., E-mail: mueller.143@osu.edu [Ohio State University, William G. Lowrie Department of Chemical and Biomolecular Engineering, OH (United States) and University of Queensland, School of Chemical Engineering (Australia) and Ohio State University, Materials Science and Engineering, OH (United States); Boehm, Michael W. [University of Queensland, School of Chemical Engineering (Australia); Drummond, Charles [Ohio State University, Materials Science and Engineering, OH (United States)

    2012-08-15T23:59:59.000Z

    Highlights: Black-Right-Pointing-Pointer Given a large flow rate of CRT glass {approx}10% of the panel glass stream will be leaded. Black-Right-Pointing-Pointer The supply of CRT waste glass exceeded demand in 2009. Black-Right-Pointing-Pointer Recyclers should use UV-light to detect lead oxide during the separation process. Black-Right-Pointing-Pointer Recycling market analysis techniques and results are given for CRT glass. Black-Right-Pointing-Pointer Academic initiatives and the necessary expansion of novel product markets are discussed. - Abstract: Cathode Ray Tube, CRT, waste glass recycling has plagued glass manufacturers, electronics recyclers and electronics waste policy makers for decades because the total supply of waste glass exceeds demand, and the formulations of CRT glass are ill suited for most reuse options. The solutions are to separate the undesirable components (e.g. lead oxide) in the waste and create demand for new products. Achieving this is no simple feat, however, as there are many obstacles: limited knowledge of waste glass composition; limited automation in the recycling process; transportation of recycled material; and a weak and underdeveloped market. Thus one of the main goals of this paper is to advise electronic glass recyclers on how to best manage a diverse supply of glass waste and successfully market to end users. Further, this paper offers future directions for academic and industry research. To develop the recommendations offered here, a combination of approaches were used: (1) a thorough study of historic trends in CRT glass chemistry; (2) bulk glass collection and analysis of cullet from a large-scale glass recycler; (3) conversations with industry members and a review of potential applications; and (4) evaluation of the economic viability of specific uses for recycled CRT glass. If academia and industry can solve these problems (for example by creating a database of composition organized by manufacturer and glass source) then the reuse of CRT glass can be increased.

  16. Field study of disposed solid wastes from advanced coal processes

    SciTech Connect (OSTI)

    Not Available

    1992-01-01T23:59:59.000Z

    Radian Corporation and the North Dakota Energy and Environmental Research Center (EERC) are funded to develop information to be used by private industry and government agencies for managing solid wastes produced by advanced coal combustion processes. This information will be developed by conducting several field studies on disposed wastes from these processes. Data will be collected to characterize these wastes and their interactions with the environments in which they are disposed. Three sites were selected for the field studies: Colorado Ute's fluidized bed combustion (FBC) unit in Nucla, Colorado; Ohio Edison's limestone injection multistage burner (LIMB) retrofit in Lorain, Ohio; and Freeman United's mine site in central Illinois with wastes supplied by the nearby Midwest Grain FBC unit. During the past year, field monitoring and sampling of the four landfill test cases constructed in 1989 and 1991 has continued. Option 1 of the contract was approved last year to add financing for the fifth test case at the Freeman United site. The construction of the Test Case 5 cells is scheduled to begin in November, 1992. Work during this past year has focused on obtaining data on the physical and chemical properties of the landfilled wastes, and on developing a conceptual framework for interpreting this information. Results to date indicate that hydration reactions within the landfilled wastes have had a major impact on the physical and chemical properties of the materials but these reactions largely ceased after the first year, and physical properties have changed little since then. Conditions in Colorado remained dry and no porewater samples were collected. In Ohio, hydration reactions and increases in the moisture content of the waste tied up much of the water initially infiltrating the test cells.

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

    DOE Patents [OSTI]

    Watson, John L. (Rolla, MO)

    1990-01-01T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Watson, J.L.

    1990-07-10T23:59:59.000Z

    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.

  19. Waste Energy Analysis Recovery for a Typical Food Processing Plant

    E-Print Network [OSTI]

    Miller, P. H.; Mann, L., Jr.

    1980-01-01T23:59:59.000Z

    An energy analysis made for the Joan of Arc Food Processing Plant in St. Francisville, Louisiana indicated that a significant quantity of waste heat energy was being released to the atmosphere in the forms of low quality steam and hot flue gases...

  20. Waste Heat Recovery from Refrigeration in a Meat Processing Facility

    E-Print Network [OSTI]

    Murphy, W. T.; Woods, B. E.; Gerdes, J. E.

    1980-01-01T23:59:59.000Z

    A case study is reviewed on a heat recovery system installed in a meat processing facility to preheat water for the plant hot water supply. The system utilizes waste superheat from the facility's 1,350-ton ammonia refrigeration system. The heat...

  1. Study on a regeneration process of LiCl-KCl eutectic based waste salt generated from the pyrochemical process

    SciTech Connect (OSTI)

    Eun, H.C.; Cho, Y.Z.; Choi, J.H.; Kim, J.H.; Lee, T.K.; Park, H.S.; Kim, I.T.; Park, G.I. [Nuclear Fuel Cycle Waste Treatment Research Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 3054-353 (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    A regeneration process of LiCl-KCl eutectic waste salt generated from the pyrochemical process of spent nuclear fuel has been studied. This regeneration process is composed of a chemical conversion process and a vacuum distillation process. Through the regeneration process, a high efficiency of renewable salt recovery can be obtained from the waste salt and rare earth nuclides in the waste salt can be separated as oxide or phosphate forms. Thus, the regeneration process can contribute greatly to a reduction of the waste volume and a creation of durable final waste forms. (authors)

  2. Selection of liquid-level monitoring method for the Oak Ridge National Laboratory inactive liquid low-level waste tanks, remedial investigation/feasibility study

    SciTech Connect (OSTI)

    Not Available

    1994-11-01T23:59:59.000Z

    Several of the inactive liquid low-level waste (LLLW) tanks at Oak Ridge National Laboratory contain residual wastes in liquid or solid (sludge) form or both. A plan of action has been developed to ensure that potential environmental impacts from the waste remaining in the inactive LLLW tank systems are minimized. This document describes the evaluation and selection of a methodology for monitoring the level of the liquid in inactive LLLW tanks. Criteria are established for comparison of existing level monitoring and leak testing methods; a preferred method is selected and a decision methodology for monitoring the level of the liquid in the tanks is presented for implementation. The methodology selected can be used to continuously monitor the tanks pending disposition of the wastes for treatment and disposal. Tanks that are empty, are scheduled to be emptied in the near future, or have liquid contents that are very low risk to the environment were not considered to be candidates for installing level monitoring. Tanks requiring new monitoring equipment were provided with conductivity probes; tanks with existing level monitoring instrumentation were not modified. The resulting data will be analyzed to determine inactive LLLW tank liquid level trends as a function of time.

  3. Mercury Reduction and Removal from High Level Waste at the Defense Waste Processing Facility - 12511

    SciTech Connect (OSTI)

    Behrouzi, Aria [Savannah River Remediation, LLC (United States); Zamecnik, Jack [Savannah River National Laboratory, Aiken, South Carolina, 29808 (United States)

    2012-07-01T23:59:59.000Z

    The Defense Waste Processing Facility processes legacy nuclear waste generated at the Savannah River Site during production of enriched uranium and plutonium required by the Cold War. The nuclear waste is first treated via a complex sequence of controlled chemical reactions and then vitrified into a borosilicate glass form and poured into stainless steel canisters. Converting the nuclear waste into borosilicate glass is a safe, effective way to reduce the volume of the waste and stabilize the radionuclides. One of the constituents in the nuclear waste is mercury, which is present because it served as a catalyst in the dissolution of uranium-aluminum alloy fuel rods. At high temperatures mercury is corrosive to off-gas equipment, this poses a major challenge to the overall vitrification process in separating mercury from the waste stream prior to feeding the high temperature melter. Mercury is currently removed during the chemical process via formic acid reduction followed by steam stripping, which allows elemental mercury to be evaporated with the water vapor generated during boiling. The vapors are then condensed and sent to a hold tank where mercury coalesces and is recovered in the tank's sump via gravity settling. Next, mercury is transferred from the tank sump to a purification cell where it is washed with water and nitric acid and removed from the facility. Throughout the chemical processing cell, compounds of mercury exist in the sludge, condensate, and off-gas; all of which present unique challenges. Mercury removal from sludge waste being fed to the DWPF melter is required to avoid exhausting it to the environment or any negative impacts to the Melter Off-Gas system. The mercury concentration must be reduced to a level of 0.8 wt% or less before being introduced to the melter. Even though this is being successfully accomplished, the material balances accounting for incoming and collected mercury are not equal. In addition, mercury has not been effectively purified and collected in the Mercury Purification Cell (MPC) since 2008. A significant cleaning campaign aims to bring the MPC back up to facility housekeeping standards. Two significant investigations are being undertaken to restore mercury collection. The SMECT mercury pump has been removed from the tank and will be functionally tested. Also, research is being conducted by the Savannah River National Laboratory to determine the effects of antifoam addition on the behavior of mercury. These path forward items will help us better understand what is occurring in the mercury collection system and ultimately lead to an improved DWPF production rate and mercury recovery rate. (authors)

  4. Waste Heat Boilers for Incineration Applications

    E-Print Network [OSTI]

    Ganapathy, V.

    Incineration is a widely used process for disposing of solid, liquid and gaseous wastes generated in various types of industries. In addition to destroying pollutants, energy may also be recovered from the waste gas streams in the form of steam...

  5. Processing of solid mixed waste containing radioactive and hazardous materials

    DOE Patents [OSTI]

    Gotovchikov, V.T.; Ivanov, A.V.; Filippov, E.A.

    1998-05-12T23:59:59.000Z

    Apparatus for the continuous heating and melting of a solid mixed waste bearing radioactive and hazardous materials to form separate metallic, slag and gaseous phases for producing compact forms of the waste material to facilitate disposal includes a copper split water-cooled (cold) crucible as a reaction vessel for receiving the waste material. The waste material is heated by means of the combination of a plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter. 6 figs.

  6. Processing of solid mixed waste containing radioactive and hazardous materials

    DOE Patents [OSTI]

    Gotovchikov, Vitaly T. (Moscow, RU); Ivanov, Alexander V. (Moscow, RU); Filippov, Eugene A. (Moscow, RU)

    1998-05-12T23:59:59.000Z

    Apparatus for the continuous heating and melting of a solid mixed waste bearing radioactive and hazardous materials to form separate metallic, slag and gaseous phases for producing compact forms of the waste material to facilitate disposal includes a copper split water-cooled (cold) crucible as a reaction vessel for receiving the waste material. The waste material is heated by means of the combination of a plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter.

  7. Separation of Metal Ions from Liquid Waste Streams

    SciTech Connect (OSTI)

    Glasgow, D. G.; Kennel, E. B.

    2004-12-01T23:59:59.000Z

    A unique mechanism was verified for removing uranium from continuously flowing aqueous solutions on a carbon nanofiber electrode with a bias voltage of -0.9 volts (dc versus Ag/AgC1). Uranium concentration was reduced from 100 ppm in the inlet feed to below 1 ppm in a single pass. Cell sizes of 1 cm, 2 inch and 4 inch evaluated during this program were all found to electrosorb uranium from an aqueous stream. The 4 inch cell performed well at uranium concentrations of 1000 ppm. Normally, ordinary electrolysis is not an option for removing uranyl ions because the electrodeposition potential is higher than the dissociation voltage of water. Thus, the ability to electrosorb uranium with greater than 99% effectiveness is a surprising result. In addition, the process was found to be reversible, so that the uranium can be released in a highly concentrated form. In addition to verifying the effectiveness of the system on bench top scale, a regeneration protocol was developed, consisting of passing a 0.1 M KNO{sub3}, solution at a pH of 2.0 and an applied potential of +1.0 V (dc versus Ag/AgC1) which resulted in a measured regeneration of 70% of the electrosorbed uranium. Other experiments studied the effect of pH on electrosorption and desorption, establishing a range of pH for both processes. Finally, it was found that, for an inlet solution of 100 ppm, the carbon nanofiber electrodes were able to electrosorb an amount of uranium in excess of 60% of the electrode mass.

  8. The Hybrid Treatment Process for mixed radioactive and hazardous waste treatment

    SciTech Connect (OSTI)

    Ross, W.A.; Kindle, C.H.

    1992-06-01T23:59:59.000Z

    This paper describes a new process for treating mixed hazardous and radioactive waste, commonly called mixed waste. The process is called the Hybrid Treatment Process (HTP), so named because it is built on the 20 years of experience with vitrification of wastes in melters, and the 12 years of experience with treatment of wastes by the in situ vitrification (ISV) process. It also uses techniques from several additional technologies. Mixed wastes are being generated by both the US Department of Energy (DOE) and by commercial sources. The wastes are those that contain both a hazardous waste regulated under the US Environmental Protection Agency's (EPA) Resource, Conservation, and Recovery Act (RCRA) regulations and a radioactive waste with source, special nuclear, or byproduct materials. The dual regulation of the wastes increases the complexity of the treatment, handling, and storage of the waste. The DOE is the largest holder and generator of mixed waste. Its mixed wastes are classified as either high-level, transuranic (TRU), or low-level waste (LLW). High-level mixed wastes will be treated in vitrification plants. Transuranic wastes may be disposed of without treatment by obtaining a no-migration variance from the EPA. Lowlevel wastes, however, will require treatment, but treatment systems with sufficient capacity are not yet available to DOE. Various facilities are being proposed for the treatment of low-level waste. The concept described in this paper represents one option for establishing that treatment capacity.

  9. The Hybrid Treatment Process for mixed radioactive and hazardous waste treatment

    SciTech Connect (OSTI)

    Ross, W.A.; Kindle, C.H.

    1992-06-01T23:59:59.000Z

    This paper describes a new process for treating mixed hazardous and radioactive waste, commonly called mixed waste. The process is called the Hybrid Treatment Process (HTP), so named because it is built on the 20 years of experience with vitrification of wastes in melters, and the 12 years of experience with treatment of wastes by the in situ vitrification (ISV) process. It also uses techniques from several additional technologies. Mixed wastes are being generated by both the US Department of Energy (DOE) and by commercial sources. The wastes are those that contain both a hazardous waste regulated under the US Environmental Protection Agency`s (EPA) Resource, Conservation, and Recovery Act (RCRA) regulations and a radioactive waste with source, special nuclear, or byproduct materials. The dual regulation of the wastes increases the complexity of the treatment, handling, and storage of the waste. The DOE is the largest holder and generator of mixed waste. Its mixed wastes are classified as either high-level, transuranic (TRU), or low-level waste (LLW). High-level mixed wastes will be treated in vitrification plants. Transuranic wastes may be disposed of without treatment by obtaining a no-migration variance from the EPA. Lowlevel wastes, however, will require treatment, but treatment systems with sufficient capacity are not yet available to DOE. Various facilities are being proposed for the treatment of low-level waste. The concept described in this paper represents one option for establishing that treatment capacity.

  10. Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario Hydro`s Bruce Nuclear Generating Station `A`

    SciTech Connect (OSTI)

    Day, J.E.; Baker, R.L.

    1995-05-01T23:59:59.000Z

    Ontario Hydro at the Bruce Nuclear Generating Station `A` has undertaken a program to render the station`s liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants to satisfy regulatory requirements for emissions. The system will remove radionuclide and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology.

  11. Environmental assessment for liquid waste treatment at the Nevada Test Site, Nye County, Nevada

    SciTech Connect (OSTI)

    NONE

    1997-01-01T23:59:59.000Z

    This environmental assessment (EA) examines the potential impacts to the environment from treatment of low-level radioactive liquid and low-level mixed liquid and semi-solid wastes generated at the Nevada Test Site (NTS). The potential impacts of the proposed action and alternative actions are discussed herein in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended in Title 42 U.S.C. (4321), and the US Department of Energy (DOE) policies and procedures set forth in Title 10 Code of Federal Regulations (CFR) Part 1021 and DOE Order 451.1, ``NEPA Compliance Program.`` The potential environmental impacts of the proposed action, construction and operation of a centralized liquid waste treatment facility, were addressed in the Final Environmental Impact Statement for the Nevada Test Site and Off-Site Locations in the State of Nevada. However, DOE is reevaluating the need for a centralized facility and is considering other alternative treatment options. This EA retains a centralized treatment facility as the proposed action but also considers other feasible alternatives.

  12. Process modeling of hydrogen production from municipal solid waste

    SciTech Connect (OSTI)

    Thorsness, C.B.

    1995-01-01T23:59:59.000Z

    The ASPEN PLUS commercial simulation software has been used to develop a process model for a conceptual process to convert municipal solid waste (MSW) to hydrogen. The process consists of hydrothermal treatment of the MSW in water to create a slurry suitable as feedstock for an oxygen blown Texaco gasifier. A method of reducing the complicated MSW feed material to a manageable set of components is outlined along with a framework for modeling the stoichiometric changes associated with the hydrothermal treatment process. Model results indicate that 0.672 kmol/s of hydrogen can be produced from the processing of 30 kg/s (2600 tonne/day) of raw MSW. A number of variations on the basic processing parameters are explored and indicate that there is a clear incentive to reduce the inert fraction in the processed slurry feed and that cofeeding a low value heavy oil may be economically attractive.

  13. Repackaging of High Fissile TRU Waste at the Transuranic Waste Processing Center - 13240

    SciTech Connect (OSTI)

    Oakley, Brian; Heacker, Fred [WAI, TRU Waste Processing Center, 100 WIPP Road Lenoir City, TN 37771 (United States)] [WAI, TRU Waste Processing Center, 100 WIPP Road Lenoir City, TN 37771 (United States); McMillan, Bill [DOE, Oak Ridge Operations, Bldg. 2714, Oak Ridge, TN 37830 (United States)] [DOE, Oak Ridge Operations, Bldg. 2714, Oak Ridge, TN 37830 (United States)

    2013-07-01T23:59:59.000Z

    Twenty-six drums of high fissile transuranic (TRU) waste from Oak Ridge National Laboratory (ORNL) operations were declared waste in the mid-1980's and placed in storage with the legacy TRU waste inventory for future treatment and disposal at the Waste Isolation Pilot Plant (WIPP). Repackaging and treatment of the waste at the TRU Waste Packaging Center (TWPC) will require the installation of additional equipment and capabilities to address the hazards for handling and repackaging the waste compared to typical Contact Handled (CH) TRU waste that is processed at the TWPC, including potential hydrogen accumulation in legacy 6M/2R packaging configurations, potential presence of reactive plutonium hydrides, and significant low energy gamma radiation dose rates. All of the waste is anticipated to be repackaged at the TWPC and certified for disposal at WIPP. The waste is currently packaged in multiple layers of containers which presents additional challenges for repackaging activities due to the potential for the accumulation of hydrogen gas in the container headspace in quantities than could exceed the Lower Flammability Limit (LFL). The outer container for each waste package is a stainless steel 0.21 m{sup 3} (55-gal) drum which contains either a 0.04 m{sup 3} or 0.06 m{sup 3} (10-gal or 15-gal) 6M drum. The inner 2R container in each 6M drum is ?12 cm (5 in) outside diameter x 30-36 cm (12-14 in) long and is considered to be a > 4 liter sealed container relative to TRU waste packaging criteria. Inside the 2R containers are multiple configurations of food pack cans, pipe nipples, and welded capsules. The waste contains significant quantities of high burn-up plutonium oxides and metals with a heavy weight percentage of higher atomic mass isotopes and the subsequent in-growth of significant quantities of americium. Significant low energy gamma radiation is expected to be present due to the americium in-growth. Radiation dose rates on inner containers are estimated to be 1-3 mSv/hr (100-300 mrem/hr) with an unshielded dose rate on the waste itself of over 10 mSv/hr (1 rem/hr). Additional equipment to be installed at the TWPC will include a new perma-con enclosure and a shielded/inert glovebox in the process building to repackage and stabilize the waste. All of the waste will be repackaged into Standard Pipe Overpacks. Most of the waste (21 of the 26 drums) is expected to be repackaged at the food-pack can level (i.e. the food-pack cans will not be opened). Five of the incoming waste containers are expected to be repackaged at the primary waste level. Three of the containers exceed the 200 gram Pu-239 Fissile Gram Equivalent (FGE) limit for the Standard Pipe Overpack. These three containers will be repackaged down to the primary waste level and divided into eight Standard Pipe Overpacks for shipment to WIPP. Two containers must be stabilized to eliminate any reactive plutonium hydrides that may be present. These containers will be opened in the inert, shielded glovebox, and the remaining corroded plutonium metal converted to a stable oxide form by using a 600 deg. C tube furnace with controlled oxygen feed in a helium carrier gas. The stabilized waste will then be packaged into two Standard Pipe Overpacks. Design and build out activities for the additional repackaging capabilities at the TWPC are scheduled to begin in Fiscal Year 2013 with repackaging, stabilization, and certification activities scheduled to begin in Fiscal Year 2014. Following repackaging and stabilization activities, the Standard Pipe Overpacks will be certified for disposal at WIPP utilizing Non-Destructive Examination (NDE) to verify the absence of prohibited items and Non-Destructive Assay (NDA) to verify the isotopic content under the TWPC WIPP certification program implemented by the Central Characterization Project (CCP). (authors)

  14. Implementation plan for liquid low-level radioactive waste systems under the FFA for fiscal years 1996 and 1997 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    NONE

    1996-06-01T23:59:59.000Z

    This document is the fourth annual revision of the plans and schedules for implementing the Federal Facility Agreement (FFA) compliance program, originally submitted in 1992 as ES/ER-17&D1, Federal Facility Agreement Plans and Schedules for Liquid Low-Level Radioactive Waste Tank Systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee. This document summarizes the progress that has been made to date implementing the plans and schedules for meeting the FFA commitments for the Liquid Low-Level Waste (LLLW) System at Oak Ridge National Laboratory (ORNL). In addition, this document lists FFA activities planned for FY 1997. Information presented in this document provides a comprehensive summary to facilitate understanding of the FFA compliance program for LLLW tank systems and to present plans and schedules associated with remediation, through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) process, of LLLW tank systems that have been removed from service.

  15. Development of US Navy Shipboard Systems for solid and liquid waste thermal treatment. Report for July 1995-April 1996

    SciTech Connect (OSTI)

    Gullet, B.K.

    1996-07-01T23:59:59.000Z

    The paper describes the U.S. Navy`s shipboard environmental challenges and a few of its research programs for meeting its needs for solid and liquid waste treatment. This objective is particularly important in environmentally sensitive areas, such as the Mediterranean Sea, where fleet deployment time is significant. Prohibitions on ocean dumping and anticipated requirements on effluent discharge quality have led the Navy to continue the research, development, and demonstration of shipboard systems to treat their unpreventable wastes. For solid, non-hazardous wastes, post-minimization efforts are geared toward long-term development of systems to thermally pyrolyze and oxidize the wastes into significantly reduced volume and weight.

  16. Treatment Options for Liquid Radioactive Waste. Factors Important for Selecting of Treatment Methods

    SciTech Connect (OSTI)

    Dziewinski, J.J.

    1998-09-28T23:59:59.000Z

    The cleanup of liquid streams contaminated with radionuclides is obtained by the selection or a combination of a number of physical and chemical separations, processes or unit operations. Among those are: Chemical treatment; Evaporation; Ion exchange and sorption; Physical separation; Electrodialysis; Osmosis; Electrocoagulation/electroflotation; Biotechnological processes; and Solvent extraction.

  17. THE USE OF POLYMERS IN RADIOACTIVE WASTE PROCESSING SYSTEMS

    SciTech Connect (OSTI)

    Skidmore, E.; Fondeur, F.

    2013-04-15T23:59:59.000Z

    The Savannah River Site (SRS), one of the largest U.S. Department of Energy (DOE) sites, has operated since the early 1950s. The early mission of the site was to produce critical nuclear materials for national defense. Many facilities have been constructed at the SRS over the years to process, stabilize and/or store radioactive waste and related materials. The primary materials of construction used in such facilities are inorganic (metals, concrete), but polymeric materials are inevitably used in various applications. The effects of aging, radiation, chemicals, heat and other environmental variables must therefore be understood to maximize service life of polymeric components. In particular, the potential for dose rate effects and synergistic effects on polymeric materials in multivariable environments can complicate compatibility reviews and life predictions. The selection and performance of polymeric materials in radioactive waste processing systems at the SRS are discussed.

  18. Multi-discipline Waste Acceptance Process at the Nevada National Security Site - 13573

    SciTech Connect (OSTI)

    Carilli, Jhon T. [US Department Of Energy, Nevada Site Office, P. O. Box 98518, Las Vegas, Nevada 89193-8518 (United States)] [US Department Of Energy, Nevada Site Office, P. O. Box 98518, Las Vegas, Nevada 89193-8518 (United States); Krenzien, Susan K. [Navarro-Intera, LLC, P. O. Box 98952, Las Vegas, Nevada 89193-8952 (United States)] [Navarro-Intera, LLC, P. O. Box 98952, Las Vegas, Nevada 89193-8952 (United States)

    2013-07-01T23:59:59.000Z

    The Nevada National Security Site low-level radioactive waste disposal facility acceptance process requires multiple disciplines to ensure the protection of workers, the public, and the environment. These disciplines, which include waste acceptance, nuclear criticality, safety, permitting, operations, and performance assessment, combine into the overall waste acceptance process to assess low-level radioactive waste streams for disposal at the Area 5 Radioactive Waste Management Site. Four waste streams recently highlighted the integration of these disciplines: the Oak Ridge Radioisotope Thermoelectric Generators and Consolidated Edison Uranium Solidification Project material, West Valley Melter, and classified waste. (authors)

  19. The Radioactive Liquid Waste Treatment Facility Replacement Project at Los Alamos National Laboratory

    Office of Environmental Management (EM)

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently AskedEnergyIssuesEnergy SolarRadioactive Liquid Waste Treatment Facility

  20. STATUS OF THE DEVELOPMENT OF IN-TANK/AT-TANK SEPARATIONS TECHNOLOGIES FOR FOR HIGH-LEVEL WASTE PROCESSING FOR THE U.S. DEPARTMENT OF ENERGY

    SciTech Connect (OSTI)

    Aaron, G.; Wilmarth, B.

    2011-09-19T23:59:59.000Z

    Within the U.S. Department of Energy's (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in 'tank farms'). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are 'first-of-a-kind' and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE's specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford's Waste Treatment and Immobilization Plant (WTP) or Savannah River's Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R&D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River. This has the potential to align the salt and sludge processing life cycle, thereby reducing the Defense Waste Processing Facility (DWPF) mission by 7 years. Additionally at the Hanford site, problematic waste streams, such as high boehmite and phosphate wastes, could be treated prior to receipt by WTP and thus dramatically improve the capacity of the facility to process HLW. Treatment of boehmite by continuous sludge leaching (CSL) before receipt by WTP will dramatically reduce the process cycle time for the WTP pretreatment facility, while treatment of phosphate will significantly reduce the number of HLW borosilicate glass canisters produced at the WTP. These and other promising technologies will be discussed.

  1. Process for hydrocracking carbonaceous material in liquid carrier

    DOE Patents [OSTI]

    Duncan, Dennis A. (Downers Grove, IL)

    1980-01-01T23:59:59.000Z

    Solid carbonaceous material is hydrocracked to provide aliphatic and aromatic hydrocarbons for use as gaseous and liquid fuels or chemical feed stock. Particulate carbonaceous material such as coal in slurry with recycled product oil is preheated in liquid state to a temperature of 600.degree.-1200.degree. F. in the presence of hydrogen gas. The product oil acts as a sorbing agent for the agglomerating bitumins to minimize caking within the process. In the hydrocracking reactor, the slurry of oil and carbonaceous particles is heated within a tubular passageway to vaporize the oil and form a gas-solid mixture which is further heated to a hydropyrolysis temperature in excess of 1200.degree. F. The gas-solid mixture is quenched by contact with additional oil to condense normally liquid hydrocarbons for separation from the gases. A fraction of the hydrocarbon liquid product is recycled for quenching and slurrying with the carbonaceous feed. Hydrogen is recovered from the gas for recycle and additional hydrogen is produced by gasification of residual char.

  2. Development of Polymeric Waste Forms for the Encapsulation of Toxic Wastes Using an Emulsion-Encapsulation Based Process

    SciTech Connect (OSTI)

    Evans, R.; Quach, A.; Birnie, D. P.; Saez, A. E.; Ela, W. P.; Zeliniski, B. J. J.; Xia, G.; Smith, H.

    2003-01-01T23:59:59.000Z

    Developed technologies in vitrification, cement, and polymeric materials manufactured using flammable organic solvents have been used to encapsulate solid wastes, including low-level radioactive materials, but are impractical for high salt-content waste streams (Maio, 1998). In this work, we investigate an emulsification process for producing an aqueous-based polymeric waste form as a preliminary step towards fabricating hybrid organic/inorganic polyceram matrices. The material developed incorporates epoxy resin and polystyrene-butadiene (PSB) latex to produce a waste form that is non-flammable, light weight, of relatively low cost, and that can be loaded to a relatively high weight content of waste materials. Sodium nitrate was used as a model for the salt waste. Small-scale samples were manufactured and analyzed using leach tests designed to measure the diffusion coefficient and leachability index for the fastest diffusing species in the waste form, the salt ions. The microstructure and composition of the samples were probed using SEM/EDS techniques. The results show that some portion of the salt migrates towards the exterior surfaces of the waste forms during the curing process. A portion of the salt in the interior of the sample is contained in polymer corpuscles or sacs. These sacs are embedded in a polymer matrix phase that contains fine, well-dispersed salt crystals. The diffusion behavior observed in sections of the waste forms indicates that samples prepared using this emulsion process meet or exceed the leachability criteria suggested for low level radioactivity waste forms.

  3. Implementation Plan for Liquid Low-Level Radioactive Waste tank systems at Oak Ridge National Laboratory under the Federal Facility Agreement, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    This document summarizes the progress that has been made to date in implementing the plans and schedules for meeting the Federal Facility Agreement (FFA) commitments for the Liquid Low-Level Waste (LLLW) System at Oak Ridge National Laboratory (ORNL). These commitments were initially submitted in ES/ER-17&Dl, Federal Facility Agreement Plans and Schedules for Liquid Low-Level Radioactive Waste Tank Systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee. Information presented in this document provides a comprehensive summary to facilitate understanding of the FFA compliance program for LLLW tank systems and to present plans and schedules associated with remediation, through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) process, of LLLW tank systems that have been removed from service. ORNL has a comprehensive program underway to upgrade the LLLW system as necessary to meet the FFA requirements. The tank systems that are removed from service are being investigated and remediated through the CERCLA process. Waste and risk characterizations have been submitted. Additional data will be prepared and submitted to EPA/TDEC as tanks are taken out of service and as required by the remedial investigation/feasibility study (RI/FS) process. The plans and schedules for implementing the FFA compliance program that were submitted in ES/ER-17&Dl, Federal Facility Agreement Plans and Schedules for Liquid Low-Level Radioactive Waste tanks Systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee, are updated in this document. Chapter 1 provides general background information and philosophies that lead to the plans and schedules that appear in Chaps. 2 through 5.

  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)

    2013-07-01T23:59:59.000Z

    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. NGNP Process Heat Utilization: Liquid Metal Phase Change Heat Exchanger

    SciTech Connect (OSTI)

    Piyush Sabharwall; Mike Patterson; Vivek Utgikar; Fred Gunnerson

    2008-09-01T23:59:59.000Z

    One key long-standing issue that must be overcome to fully realize the successful growth of nuclear power is to determine other benefits of nuclear energy apart from meeting the electricity demands. The Next Generation Nuclear Plant (NGNP) will most likely be producing electricity and heat for the production of hydrogen and/or oil retrieval from oil sands and oil shale to help in our national pursuit of energy independence. For nuclear process heat to be utilized, intermediate heat exchange is required to transfer heat from the NGNP to the hydrogen plant or oil recovery field in the most efficient way possible. Development of nuclear reactor - process heat technology has intensified the interest in liquid metals as heat transfer media because of their ideal transport properties. Liquid metal heat exchangers are not new in practical applications. An important rational for considering liquid metals is the potential convective heat transfer is among the highest known. Thus explains the interest in liquid metals as coolant for intermediate heat exchange from NGNP. For process heat it is desired that, intermediate heat exchangers (IHX) transfer heat from the NGNP in the most efficient way possible. The production of electric power at higher efficiency via the Brayton Cycle, and hydrogen production, requires both heat at higher temperatures and high effectiveness compact heat exchangers to transfer heat to either the power or process cycle. Compact heat exchangers maximize the heat transfer surface area per volume of heat exchanger; this has the benefit of reducing heat exchanger size and heat losses. High temperature IHX design requirements are governed in part by the allowable temperature drop between the outlet and inlet of the NGNP. In order to improve the characteristics of heat transfer, liquid metal phase change heat exchangers may be more effective and efficient. This paper explores the overall heat transfer characteristics and pressure drop of the phase change heat exchanger with Na as the heat exchanger coolant. In order to design a very efficient and effective heat exchanger one must optimize the design such that we have a high heat transfer and a lower pressure drop, but there is always a trade-off between them. Based on NGNP operational parameters, a heat exchanger analysis with the sodium phase change will be presented to show that the heat exchanger has the potential for highly effective heat transfer, within a small volume at reasonable cost.

  6. Overview of Fiscal Year 2002 Research and Development for Savannah River Site's Salt Waste Processing Facility

    SciTech Connect (OSTI)

    H. D. Harmon, R. Leugemors, PNNL; S. Fink, M. Thompson, D. Walker, WSRC; P. Suggs, W. D. Clark, Jr

    2003-02-26T23:59:59.000Z

    The Department of Energy's (DOE) Savannah River Site (SRS) high-level waste program is responsible for storage, treatment, and immobilization of high-level waste for disposal. The Salt Processing Program (SPP) is the salt (soluble) waste treatment portion of the SRS high-level waste effort. The overall SPP encompasses the selection, design, construction and operation of treatment technologies to prepare the salt waste feed material for the site's grout facility (Saltstone) and vitrification facility (Defense Waste Processing Facility). Major constituents that must be removed from the salt waste and sent as feed to Defense Waste Processing Facility include actinides, strontium, cesium, and entrained sludge. In fiscal year 2002 (FY02), research and development (R&D) on the actinide and strontium removal and Caustic-Side Solvent Extraction (CSSX) processes transitioned from technology development for baseline process selection to providing input for conceptual design of the Salt Waste Processing Facility. The SPP R&D focused on advancing the technical maturity, risk reduction, engineering development, and design support for DOE's engineering, procurement, and construction (EPC) contractors for the Salt Waste Processing Facility. Thus, R&D in FY02 addressed the areas of actual waste performance, process chemistry, engineering tests of equipment, and chemical and physical properties relevant to safety. All of the testing, studies, and reports were summarized and provided to the DOE to support the Salt Waste Processing Facility, which began conceptual design in September 2002.

  7. Solid Waste Processing Center Primary Opening Cells Systems, Equipment and Tools

    SciTech Connect (OSTI)

    Bailey, Sharon A.; Baker, Carl P.; Mullen, O Dennis; Valdez, Patrick LJ

    2006-04-17T23:59:59.000Z

    This document addresses the remote systems and design integration aspects of the development of the Solid Waste Processing Center (SWPC), a facility to remotely open, sort, size reduce, and repackage mixed low-level waste (MLLW) and transuranic (TRU)/TRU mixed waste that is either contact-handled (CH) waste in large containers or remote-handled (RH) waste in various-sized packages.

  8. Audit Report on "Waste Processing and Recovery Act Acceleration Efforts for Contact-Handled Transuranic Waste at the Hanford Site"

    SciTech Connect (OSTI)

    None

    2010-05-01T23:59:59.000Z

    The Department of Energy's Office of Environmental Management's (EM), Richland Operations Office (Richland), is responsible for disposing of the Hanford Site's (Hanford) transuranic (TRU) waste, including nearly 12,000 cubic meters of radioactive contact-handled TRU wastes. Prior to disposing of this waste at the Department's Waste Isolation Pilot Plant (WIPP), Richland must certify that it meets WIPP's waste acceptance criteria. To be certified, the waste must be characterized, screened for prohibited items, treated (if necessary) and placed into a satisfactory disposal container. In a February 2008 amendment to an existing Record of Decision (Decision), the Department announced its plan to ship up to 8,764 cubic meters of contact-handled TRU waste from Hanford and other waste generator sites to the Advanced Mixed Waste Treatment Project (AMWTP) at Idaho's National Laboratory (INL) for processing and certification prior to disposal at WIPP. The Department decided to maximize the use of the AMWTP's automated waste processing capabilities to compact and, thereby, reduce the volume of contact-handled TRU waste. Compaction reduces the number of shipments and permits WIPP to more efficiently use its limited TRU waste disposal capacity. The Decision noted that the use of AMWTP would avoid the time and expense of establishing a processing capability at other sites. In May 2009, EM allocated $229 million of American Recovery and Reinvestment Act of 2009 (Recovery Act) funds to support Hanford's Solid Waste Program, including Hanford's contact-handled TRU waste. Besides providing jobs, these funds were intended to accelerate cleanup in the short term. We initiated this audit to determine whether the Department was effectively using Recovery Act funds to accelerate processing of Hanford's contact-handled TRU waste. Relying on the availability of Recovery Act funds, the Department changed course and approved an alternative plan that could increase costs by about $25 million by processing Hanford TRU-waste on-site rather than at AMWTP. Further, under the newly adopted alternative approach, the Department would fail to achieve the previously anticipated reductions in volume associated with the use of existing AMWTP waste compaction capabilities.

  9. Final Treatment Center Project for Liquid and Wet Radioactive Waste in Slovakia

    SciTech Connect (OSTI)

    Kravarik, K.; Stubna, M.; Pekar, A.; Krajc, T.; Zatkulak, M.; Holicka, Z. [VUJE, Inc., Okruzna 5, 918 64 Trnava (Slovakia); Slezak, M. [SE - VYZ, 919 31 Jaslovske Bohunice (Slovakia)

    2006-07-01T23:59:59.000Z

    The Final Treatment Center (FTC) for Mochovce nuclear power plant (NPP) is designed for treatment and final conditioning of radioactive liquid and wet waste produced from plant operation. Mochovce NNP uses a Russian VVER-440 type reactor. Treated wastes comprise radioactive concentrates, spent resin and sludge. VUJE Inc. as an experienced company in field of treatment of radioactive waste in Slovakia has been chosen as main contractor for technological part of FTC. This paper describes the capacity, flow chart, overall waste flow and parameters of the main components in the FTC. The initial project was submitted for approval to the Slovak Electric plc. in 2003. The design and manufacture of main components were performed in 2004 and 2005. FTC construction work started early in 2004. Initial non-radioactive testing of the system is planned for summer 2006 and then radioactive tests are to be followed. A one-year trial operation of facility is planned for completion in 2007. SE - VYZ will be operates the FTC during trial operation and after its completion. SE - VYZ is subsidiary company of Slovak Electric plc. and it is responsible for treatment with radioactive waste and spent fuel in the Slovak republic. SE - VYZ has, besides of other significant experience with operation of Jaslovske Bohunice Treatment Centre. The overall capacity of the FTC is 870 m{sup 3}/year of concentrates and 40 m{sup 3}/year of spent resin and sludge. Bituminization and cementation were provided as main technologies for treatment of these wastes. Treatment of concentrate is performed by bituminization. Concentrate and bitumen are metered into a thin film evaporator with rotating wiping blades. Surplus water is evaporated and concentrate salts are embedded in bitumen. Bitumen product is discharged into 200 l steel drums. Spent resin and sludge are decanted, dried and mixed with bitumen. These mixtures are also discharged into 200 l steel drums. Drums are moved along bituminization line on a roller conveyor. After the drums cool, they are capped and removed from the conveyor and placed in a storage hall. Drums with bitumen product are loaded into Fiber Reinforced Concrete containers (FRC) and grouted with cement. Cement grout is prepared from mixture of cement, additive and radioactive concentrates. By formulating the cement grout with evaporator concentrates the maximum radioactivity is fixed in cement matrix and volume of final waste product is minimized. A batch mixer with rotating blades is used produce the cement grout. FRCs loaded with bitumen drums are placed on roller conveyor and moved along the cementation line. Grouted FRCs are stored in the expedition hall for 28 days of curing and then transported to final disposal. After placed in operation the FTC provides treatment for all liquid and wet LLW produced from the operation of the Mochovce NPP. The final product of the FTC is a FRC loaded with 7 drums of waste fixed in bitumen and the space between the drums is grouted with cement. This container meets all limits for final disposal in the National Radioactive Waste Repository at Mochovce. (authors)

  10. Hanford Facility dangerous waste permit application, liquid effluent retention facility and 200 area effluent treatment facility

    SciTech Connect (OSTI)

    Coenenberg, J.G.

    1997-08-15T23:59:59.000Z

    The Hanford Facility Dangerous Waste Permit Application is considered to 10 be a single application organized into a General Information Portion (document 11 number DOE/RL-91-28) and a Unit-Specific Portion. The scope of the 12 Unit-Specific Portion is limited to Part B permit application documentation 13 submitted for individual, `operating` treatment, storage, and/or disposal 14 units, such as the Liquid Effluent Retention Facility and 200 Area Effluent 15 Treatment Facility (this document, DOE/RL-97-03). 16 17 Both the General Information and Unit-Specific portions of the Hanford 18 Facility Dangerous Waste Permit Application address the content of the Part B 19 permit application guidance prepared by the Washington State Department of 20 Ecology (Ecology 1987 and 1996) and the U.S. Environmental Protection Agency 21 (40 Code of Federal Regulations 270), with additional information needs 22 defined by the Hazardous and Solid Waste Amendments and revisions of 23 Washington Administrative Code 173-303. For ease of reference, the Washington 24 State Department of Ecology alpha-numeric section identifiers from the permit 25 application guidance documentation (Ecology 1996) follow, in brackets, the 26 chapter headings and subheadings. A checklist indicating where information is 27 contained in the Liquid Effluent Retention Facility and 200 Area Effluent 28 Treatment Facility permit application documentation, in relation to the 29 Washington State Department of Ecology guidance, is located in the Contents 30 Section. 31 32 Documentation contained in the General Information Portion is broader in 33 nature and could be used by multiple treatment, storage, and/or disposal units 34 (e.g., the glossary provided in the General Information Portion). Wherever 35 appropriate, the Liquid Effluent Retention Facility and 200 Area Effluent 36 Treatment Facility permit application documentation makes cross-reference to 37 the General Information Portion, rather than duplicating text. 38 39 Information provided in this Liquid Effluent Retention Facility and 40 200 Area Effluent Treatment Facility permit application documentation is 41 current as of June 1, 1997.

  11. Towards model-based control of a steam Rankine process for engine waste heat recovery

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Towards model-based control of a steam Rankine process for engine waste heat recovery Johan Peralez a critical role in enabling good per- formance of Rankine processes for waste heat recovery from prime movers. INTRODUCTION In the last few years, engine waste heat recovery (WHR) systems based on the Rankine thermodynamic

  12. Sampling and analysis of radioactive liquid wastes and sludges in the Melton Valley and evaporator facility storage tanks at ORNL

    SciTech Connect (OSTI)

    Sears, M.B.; Botts, J.L.; Ceo, R.N.; Ferrada, J.J.; Griest, W.H.; Keller, J.M.; Schenley, R.L.

    1990-09-01T23:59:59.000Z

    The sampling and analysis of the radioactive liquid wastes and sludges in the Melton Valley Storage Tanks (MVSTs), as well as two of the evaporator service facility storage tanks at ORNL, are described. Aqueous samples of the supernatant liquid and composite samples of the sludges were analyzed for major constituents, radionuclides, total organic carbon, and metals listed as hazardous under the Resource Conservation and Recovery Act (RCRA). Liquid samples from five tanks and sludge samples from three tanks were analyzed for organic compounds on the Environmental Protection Agency (EPA) Target Compound List. Estimates were made of the inventory of liquid and sludge phases in the tanks. Descriptions of the sampling and analytical activities and tabulations of the results are included. The report provides data in support of the design of the proposed Waste Handling and Packaging Plant, the Liquid Low-Level Waste Solidification Project, and research and development activities (R D) activities in developing waste management alternatives. 7 refs., 8 figs., 16 tabs.

  13. Modeling Coupled Processes in Clay Formations for Radioactive Waste Disposal

    E-Print Network [OSTI]

    Liu, Hui-Hai

    2010-01-01T23:59:59.000Z

    sorption including waste heat, hyperalkaline solutions frome.g. , heat production from the decay of the waste, re-waste packages along the tunnels, to achieve a distributed heat

  14. Tank 42 sludge-only process development for the Defense Waste Processing Facility (DWPF)

    SciTech Connect (OSTI)

    Lambert, D.P.

    2000-03-22T23:59:59.000Z

    Defense Waste Processing Facility (DWPF) requested the development of a sludge-only process for Tank 42 sludge since at the current processing rate, the Tank 51 sludge has been projected to be depleted as early as August 1998. Testing was completed using a non-radioactive Tank 42 sludge simulant. The testing was completed under a range of operating conditions, including worst case conditions, to develop the processing conditions for radioactive Tank 42 sludge. The existing Tank 51 sludge-only process is adequate with the exception that 10 percent additional acid is recommended during sludge receipt and adjustment tank (SRAT) processing to ensure adequate destruction of nitrite during the SRAT cycle.

  15. Process for treating waste water having low concentrations of metallic contaminants

    DOE Patents [OSTI]

    Looney, Brian B; Millings, Margaret R; Nichols, Ralph L; Payne, William L

    2014-12-16T23:59:59.000Z

    A process for treating waste water having a low level of metallic contaminants by reducing the toxicity level of metallic contaminants to an acceptable level and subsequently discharging the treated waste water into the environment without removing the treated contaminants.

  16. Process of producing liquid hydrocarbon fuels from biomass

    DOE Patents [OSTI]

    Kuester, James L. (Scottsdale, AZ)

    1987-07-07T23:59:59.000Z

    A continuous thermochemical indirect liquefaction process to convert various biomass materials into diesel-type transportation fuels which fuels are compatible with current engine designs and distribution systems comprising feeding said biomass into a circulating solid fluidized bed gasification system to produce a synthesis gas containing olefins, hydrogen and carbon monoxide and thereafter introducing the synthesis gas into a catalytic liquefaction system to convert the synthesis gas into liquid hydrocarbon fuel consisting essentially of C.sub.7 -C.sub.17 paraffinic hydrocarbons having cetane indices of 50+.

  17. Process of producing liquid hydrocarbon fuels from biomass

    DOE Patents [OSTI]

    Kuester, J.L.

    1987-07-07T23:59:59.000Z

    A continuous thermochemical indirect liquefaction process is described to convert various biomass materials into diesel-type transportation fuels which fuels are compatible with current engine designs and distribution systems comprising feeding said biomass into a circulating solid fluidized bed gasification system to produce a synthesis gas containing olefins, hydrogen and carbon monoxide and thereafter introducing the synthesis gas into a catalytic liquefaction system to convert the synthesis gas into liquid hydrocarbon fuel consisting essentially of C[sub 7]-C[sub 17] paraffinic hydrocarbons having cetane indices of 50+. 1 fig.

  18. Waste processing cost recovery at Los Alamos National Laboratory--analysis and recommendations

    SciTech Connect (OSTI)

    Booth, Steven Richard [Los Alamos National Laboratory

    2008-01-01T23:59:59.000Z

    Los Alamos National Laboratory is implementing full cost recovery for waste processing in fiscal year 2009 (FY2009), after a transition year in FY2008. Waste processing cost recovery has been implemented in various forms across the nuclear weapons complex and in corporate America. The fundamental reasoning of sending accurate price signals to waste generators is economically sound, and leads to waste minimization and reduced waste expense over time. However, Los Alamos faces significant implementation challenges because of its status as a government-owned, contractor-operated national scientific institution with a diverse suite of experimental and environmental cleanup activities, and the fact that this represents a fundamental change in how waste processing is viewed by the institution. This paper describes the issues involved during the transition to cost recovery and the ultimate selection of the business model. Of the six alternative cost recovery models evaluated, the business model chosen to be implemented in FY2009 is Recharge Plus Generators Pay Distributed Direct. Under this model, all generators who produce waste must pay a distributed direct share associated with their specific waste type to use a waste processing capability. This cost share is calculated using the distributed direct method on the fixed cost only, i.e., the fixed cost share is based on each program's forecast proportion of the total Los Alamos volume forecast of each waste type. (Fixed activities are those required to establish the waste processing capability, i.e., to make the process ready, permitted, certified, and prepared to handle the first unit ofwaste. Therefore, the fixed cost ends at the point just before waste begins 'to be processed. The activities to actually process the waste are considered variable.) The volume of waste actually sent for processing is charged a unit cost based solely on the variable cost of disposing of that waste. The total cost recovered each year is the total distributed direct shares from generators plus the unit cost times actual volumes processed.

  19. Tank Waste Remediation System tank waste pretreatment and vitrification process development testing requirements assessment

    SciTech Connect (OSTI)

    Howden, G.F.

    1994-10-24T23:59:59.000Z

    A multi-faceted study was initiated in November 1993 to provide assurance that needed testing capabilities, facilities, and support infrastructure (sampling systems, casks, transportation systems, permits, etc.) would be available when needed for process and equipment development to support pretreatment and vitrification facility design and construction schedules. This first major report provides a snapshot of the known testing needs for pretreatment, low-level waste (LLW) and high-level waste (HLW) vitrification, and documents the results of a series of preliminary studies and workshops to define the issues needing resolution by cold or hot testing. Identified in this report are more than 140 Hanford Site tank waste pretreatment and LLW/HLW vitrification technology issues that can only be resolved by testing. The report also broadly characterizes the level of testing needed to resolve each issue. A second report will provide a strategy(ies) for ensuring timely test capability. Later reports will assess the capabilities of existing facilities to support needed testing and will recommend siting of the tests together with needed facility and infrastructure upgrades or additions.

  20. Process for decontaminating radioactive liquids using a calcium cyanamide-containing composition. [Patent application

    DOE Patents [OSTI]

    Silver, G.L.

    1980-09-24T23:59:59.000Z

    The present invention provides a process for decontaminating a radioactive liquid containing a radioactive element capable of forming a hydroxide. This process includes the steps of contacting the radioactive liquid with a decontaminating composition and separating the resulting radioactive sludge from the resulting liquid. The decontaminating composition contains calcium cyanamide.

  1. Process Knowledge Summary Report for Materials and Fuels Complex Contact-Handled Transuranic Debris Waste

    SciTech Connect (OSTI)

    R. P. Grant; P. J. Crane; S. Butler; M. A. Henry

    2010-02-01T23:59:59.000Z

    This Process Knowledge Summary Report summarizes the information collected to satisfy the transportation and waste acceptance requirements for the transfer of transuranic (TRU) waste between the Materials and Fuels Complex (MFC) and the Advanced Mixed Waste Treatment Project (AMWTP). The information collected includes documentation that addresses the requirements for AMWTP and the applicable portion of their Resource Conservation and Recovery Act permits for receipt and treatment of TRU debris waste in AMWTP. This report has been prepared for contact-handled TRU debris waste generated by the Idaho National Laboratory at MFC. The TRU debris waste will be shipped to AMWTP for purposes of supercompaction. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU debris waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for waste originating from MFC.

  2. Simulation, integration, and economic analysis of gas-to-liquid processes 

    E-Print Network [OSTI]

    Bao, Buping

    2009-05-15T23:59:59.000Z

    Gas-to-liquid (GTL) process involves the chemical conversion of natural gas (or other gas sources) into synthetic crude that can be upgraded and separated into different useful hydrocarbon fractions including liquid transportation fuels. A leading...

  3. Process Knowledge Characterization of Radioactive Waste at the Classified Waste Landfill Remediation Project Sandia National Laboratories, Albuquerque, New Mexico

    SciTech Connect (OSTI)

    DOTSON,PATRICK WELLS; GALLOWAY,ROBERT B.; JOHNSON JR,CARL EDWARD

    1999-11-03T23:59:59.000Z

    This paper discusses the development and application of process knowledge (PK) to the characterization of radioactive wastes generated during the excavation of buried materials at the Sandia National Laboratories/New Mexico (SNL/NM) Classified Waste Landfill (CWLF). The CWLF, located in SNL/NM Technical Area II, is a 1.5-acre site that received nuclear weapon components and related materials from about 1950 through 1987. These materials were used in the development and testing of nuclear weapon designs. The CWLF is being remediated by the SNL/NM Environmental Restoration (ER) Project pursuant to regulations of the New Mexico Environment Department. A goal of the CWLF project is to maximize the amount of excavated materials that can be demilitarized and recycled. However, some of these materials are radioactively contaminated and, if they cannot be decontaminated, are destined to require disposal as radioactive waste. Five major radioactive waste streams have been designated on the CWLF project, including: unclassified soft radioactive waste--consists of soft, compatible trash such as paper, plastic, and plywood; unclassified solid radioactive waste--includes scrap metal, other unclassified hardware items, and soil; unclassified mixed waste--contains the same materials as unclassified soft or solid radioactive waste, but also contains one or more Resource Conservation and Recovery Act (RCRA) constituents; classified radioactive waste--consists of classified artifacts, usually weapons components, that contain only radioactive contaminants; and classified mixed waste--comprises radioactive classified material that also contains RCRA constituents. These waste streams contain a variety of radionuclides that exist both as surface contamination and as sealed sources. To characterize these wastes, the CWLF project's waste management team is relying on data obtained from direct measurement of radionuclide activity content to the maximum extent possible and, in cases where direct measurement is not technically feasible, from accumulated PK of the excavated materials.

  4. Radioactive Waste Radioactive Waste

    E-Print Network [OSTI]

    Slatton, Clint

    form · Separate liquid from solid · Radionuclide · Separate all but H3/C14 #12;#12;Radioactive Waste;Radioactive Waste H3/C14 solids Type B (non-incinerable) metal glass hazardous materials #12;#12;Radioactive#12;Radioactive Waste at UF Bldg 831 392-8400 #12;Radioactive Waste · Program is designed to

  5. Microwave-assisted chemical process for treatment of hazardous waste: Annual report

    SciTech Connect (OSTI)

    Varma, R.; Nandi, S.P.; Cleaveland, D.C.

    1987-10-01T23:59:59.000Z

    Microwave energy provides rapid in situ uniform heating and can be used to initiate chemical processes at moderate temperatures. We investigate the technical feasibility of microwave-assisted chemical processes for detoxification of liquid hazardous waste. Trichloroethylene, a major constituent of waste streams, was selected for this detoxification study. Experiments were performed to investigate the oxidative degradation of trichloroethylene over active carbons (with and without catalysts) in air streams with microwave in situ heating, and to examine the feasibility of regenerating the used carbons. This study established that trichloroethylene in a vapor stream can be adsorbed at room temperature on active carbon beds that are loaded with Cu and Cr catalysts. When the bed is heated by a microwave radiation to moderate temperatures (<400/sup 0/C) while a moist air stream is passed through it, the trichloroethylene is readily converted into less-noxious products such as HCl, CO, CO/sub 2/ and C/sub 2/H/sub 2/Cl/sub 2/. Conversion higher than 80% was observed. Furthermore, the used carbon bed can be conveniently regenerated by microwave heating while a moist-N/sub 2/ or moist-air stream is passed through the bed. 4 refs., 5 figs., 10 tabs.

  6. Electromagnetic mixed-waste processing system for asbestos decontamination

    SciTech Connect (OSTI)

    NONE

    1995-04-01T23:59:59.000Z

    The first phase of a program to develop and demonstrate a cost-effective, integrated process for remediation of asbestos-containing material that is contaminated with organics, heavy metals, and radioactive compounds was successfully completed. Laboratory scale tests were performed to demonstrate initial process viability for asbestos conversion, organics removal, and radionuclide and heavy metal removal. All success criteria for the laboratory tests were met. (1) Ohio DSI demonstrated greater than 99% asbestos conversion to amorphous solids using their commercial process. (2) KAI demonstrated 90% removal of organics from the asbestos suspension. (3) Westinghouse STC achieved the required metals removal criteria on a laboratory scale (e.g., 92% removal of uranium from solution, resin loadings of 0.6 equivalents per liter, and greater than 50% regeneration of resin in a batch test.) Using the information gained in the laboratory tests, the process was reconfigured to provide the basis for the mixed waste remediation system. An integrated process is conceptually developed, and a Phase 2 program plan is proposed to provide the bench-scale development needed in order to refine the design basis for a pilot processing system.

  7. Tank waste remediation system process engineering instruction manual

    SciTech Connect (OSTI)

    ADAMS, M.R.

    1998-11-04T23:59:59.000Z

    The purpose of the Tank Waste Remediation System (TWRS) Process Engineering Instruction Manual is to provide guidance and direction to TWRS Process Engineering staff regarding conduct of business. The objective is to establish a disciplined and consistent approach to business such that the work processes within TWRS Process Engineering are safe, high quality, disciplined, efficient, and consistent with Lockheed Martin Hanford Corporation Policies and Procedures. The sections within this manual are of two types: for compliance and for guidance. For compliance sections are intended to be followed per-the-letter until such time as they are formally changed per Section 2.0 of this manual. For guidance sections are intended to be used by the staff for guidance in the conduct of work where technical judgment and discernment are required. The guidance sections shall also be changed per Section 2.0 of this manual. The required header for each manual section is illustrated in Section 2.0, Manual Change Control procedure. It is intended that this manual be used as a training and indoctrination resource for employees of the TWRS Process Engineering organization. The manual shall be required reading for all TWRS Process Engineering staff, matrixed, and subcontracted employees.

  8. Modeling Coupled Processes in Clay Formations for Radioactive Waste Disposal

    SciTech Connect (OSTI)

    Liu, Hui-Hai; Rutqvist, Jonny; Zheng, Liange; Sonnenthal, Eric; Houseworth, Jim; Birkholzer, Jens

    2010-08-31T23:59:59.000Z

    As a result of the termination of the Yucca Mountain Project, the United States Department of Energy (DOE) has started to explore various alternative avenues for the disposition of used nuclear fuel and nuclear waste. The overall scope of the investigation includes temporary storage, transportation issues, permanent disposal, various nuclear fuel types, processing alternatives, and resulting waste streams. Although geologic disposal is not the only alternative, it is still the leading candidate for permanent disposal. The realm of geologic disposal also offers a range of geologic environments that may be considered, among those clay shale formations. Figure 1-1 presents the distribution of clay/shale formations within the USA. Clay rock/shale has been considered as potential host rock for geological disposal of high-level nuclear waste throughout the world, because of its low permeability, low diffusion coefficient, high retention capacity for radionuclides, and capability to self-seal fractures induced by tunnel excavation. For example, Callovo-Oxfordian argillites at the Bure site, France (Fouche et al., 2004), Toarcian argillites at the Tournemire site, France (Patriarche et al., 2004), Opalinus clay at the Mont Terri site, Switzerland (Meier et al., 2000), and Boom clay at Mol site, Belgium (Barnichon et al., 2005) have all been under intensive scientific investigations (at both field and laboratory scales) for understanding a variety of rock properties and their relations with flow and transport processes associated with geological disposal of nuclear waste. Clay/shale formations may be generally classified as indurated and plastic clays (Tsang et al., 2005). The latter (including Boom clay) is a softer material without high cohesion; its deformation is dominantly plastic. For both clay rocks, coupled thermal, hydrological, mechanical and chemical (THMC) processes are expected to have a significant impact on the long-term safety of a clay repository. For example, the excavation-damaged zone (EDZ) near repository tunnels can modify local permeability (resulting from induced fractures), potentially leading to less confinement capability (Tsang et al., 2005). Because of clay's swelling and shrinkage behavior (depending on whether the clay is in imbibition or drainage processes), fracture properties in the EDZ are quite dynamic and evolve over time as hydromechanical conditions change. To understand and model the coupled processes and their impact on repository performance is critical for the defensible performance assessment of a clay repository. Within the Natural Barrier System (NBS) group of the Used Fuel Disposition (UFD) Campaign at DOE's Office of Nuclear Energy, LBNL's research activities have focused on understanding and modeling such coupled processes. LBNL provided a report in this April on literature survey of studies on coupled processes in clay repositories and identification of technical issues and knowledge gaps (Tsang et al., 2010). This report will document other LBNL research activities within the natural system work package, including the development of constitutive relationships for elastic deformation of clay rock (Section 2), a THM modeling study (Section 3) and a THC modeling study (Section 4). The purpose of the THM and THC modeling studies is to demonstrate the current modeling capabilities in dealing with coupled processes in a potential clay repository. In Section 5, we discuss potential future R&D work based on the identified knowledge gaps. The linkage between these activities and related FEPs is presented in Section 6.

  9. Design and construction of the defense waste processing facility project at the Savannah River Plant

    SciTech Connect (OSTI)

    Baxter, R G

    1986-01-01T23:59:59.000Z

    The Du Pont Company is building for the Department of Energy a facility to vitrify high-level radioactive waste at the Savannah River Plant (SRP) near Aiken, South Carolina. The Defense Waste Processing Facility (DWPF) will solidify existing and future radioactive wastes by immobilizing the waste in Processing Facility (DWPF) will solidify existing and future radioactives wastes by immobilizing the waste in borosilicate glass contained in stainless steel canisters. The canisters will be sealed, decontaminated and stored, prior to emplacement in a federal repository. At the present time, engineering and design is 90% complete, construction is 25% complete, and radioactive processing in the $870 million facility is expected to begin by late 1989. This paper describes the SRP waste characteristics, the DWPF processing, building and equipment features, and construction progress of the facility.

  10. 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.

    1981-10-01T23:59:59.000Z

    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.

  11. Surveillance and maintenance plan for the inactive liquid low-level waste tanks at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Not Available

    1994-11-01T23:59:59.000Z

    ORNL has a total of 54 inactive liquid low-level waste (ILLLW) tanks. In the past, these tanks were used to contain radioactive liquid wastes from various research programs, decontamination operations, and reactor operations. The tanks have since been removed from service for various reasons; the majority were retired because of their age, some due to integrity compromises, and others because they did not meet the current standards set by the Federal Facilities Agreement (FFA). Many of the tanks contain residual radioactive liquids and/or sludges. Plans are to remediate all tanks; however, until remediation of each tank, this Surveillance and Maintenance (S&M) Plan will be used to monitor the safety and inventory containment of these tanks.

  12. Technology for Treatment of Liquid Radioactive Waste Generated during Uranium and Plutonium Chemical and Metallurgical Manufacturing in FSUE PO Mayak - 13616

    SciTech Connect (OSTI)

    Adamovich, D. [SUE MosSIA Radon, 2/14 7th Rostovsky lane, Moscow, 119121 (Russian Federation)] [SUE MosSIA Radon, 2/14 7th Rostovsky lane, Moscow, 119121 (Russian Federation); Batorshin, G.; Logunov, M.; Musalnikov, A. [FSUE 'PO Mayak', 31 av. Lenin, Ozyorsk, Chelyabinsk region, 456780 (Russian Federation)] [FSUE 'PO Mayak', 31 av. Lenin, Ozyorsk, Chelyabinsk region, 456780 (Russian Federation)

    2013-07-01T23:59:59.000Z

    Created technological scheme for treatment of liquid radioactive waste generated while uranium and plutonium chemical and metallurgical manufacturing consists of: - Liquid radioactive waste (LRW) purification from radionuclides and its transfer into category of manufacturing waste; - Concentration of suspensions containing alpha-nuclides and their further conversion to safe dry state (calcinate) and moving to long controlled storage. The following technologies are implemented in LRW treatment complex: - Settling and filtering technology for treatment of liquid intermediate-level waste (ILW) with volume about 1500m{sup 3}/year and alpha-activity from 10{sup 6} to 10{sup 8} Bq/dm{sup 3} - Membrane and sorption technology for processing of low-level waste (LLW) of radioactive drain waters with volume about 150 000 m{sup 3}/year and alpha-activity from 10{sup 3} to 10{sup 4} Bq/dm{sup 3}. Settling and filtering technology includes two stages of ILW immobilization accompanied with primary settling of radionuclides on transition metal hydroxides with the following flushing and drying of the pulp generated; secondary deep after settling of radionuclides on transition metal hydroxides with the following solid phase concentration by the method of tangential flow ultrafiltration. Besides, the installation capacity on permeate is not less than 3 m{sup 3}/h. Concentrates generated are sent to calcination on microwave drying (MW drying) unit. Membrane and sorption technology includes processing of averaged sewage flux by the method of tangential flow ultrafiltration with total capacity of installations on permeate not less than 18 m{sup 3}/h and sorption extraction of uranium from permeate on anionite. According to radionuclide contamination level purified solution refers to general industrial waste. Concentrates generated during suspension filtering are evaporated in rotary film evaporator (RFE) in order to remove excess water, thereafter they are dried on infrared heating facility. Solid concentrate produced is sent for long controlled storage. Complex of the procedures carried out makes it possible to solve problems on treatment of LRW generated while uranium and plutonium chemical and metallurgical manufacturing in Federal State Unitary Enterprise (FSUE) Mayak and cease its discharge into open water reservoirs. (authors)

  13. Conditioning of Boron-Containing Low and Intermediate Level Liquid Radioactive Waste - 12041

    SciTech Connect (OSTI)

    Gorbunova, Olga A. [SUE SIA 'Radon', Moscow (Russian Federation); Kamaeva, Tatiana S. [Vernadsky Institute of Geochemistry and Analytical Chemistry Russian Academy of Sciences, Moscow (Russian Federation)

    2012-07-01T23:59:59.000Z

    Improved cementation of low and intermediate level radioactive waste (ILW and LLW) aided by vortex electromagnetic treatment as well as silica addition was investigated. Positive effects including accelerated curing of boron-containing cement waste forms, improve end product quality, decreased product volume and reduced secondary LRW volume from equipment decontamination were established. These results established the possibility of boron-containing LRW cementation without the use of neutralizing alkaline additives that greatly increase the volume of the final product intended for long-term storage (burial). Physical (electromagnetic) treatment in a vortex mixer can change the state of LRW versus chemical treatment. By treating the liquid phase of cement solution only, instead of the whole solution, and using fine powder and nano-particles of ferric oxides instead of separable ferromagnetic cores for the activating agents the positive effect are obtained. VET for 1 to 3 minutes yields boron-containing LRW cemented products of satisfactory quality. Silica addition at 10 % by weight will accelerate curing and solidification and to decrease radionuclide leaching rates from boron-containing cement products. (authors)

  14. Defense Waste Processing Facility wasteform and canister description: Revision 2

    SciTech Connect (OSTI)

    Baxter, R.G.

    1988-12-01T23:59:59.000Z

    This document describes the reference wasteform and canister for the Defense Waste Processing Facility (DWPF). The principal changes include revised feed and glass product compositions, an estimate of glass product characteristics as a function of time after the start of vitrification, and additional data on glass leaching performance. The feed and glass product composition data are identical to that described in the DWPF Basic Data Report, Revision 90/91. The DWPF facility is located at the Savannah River Plant in Aiken, SC, and it is scheduled for construction completion during December 1989. The wasteform is borosilicate glass containing approximately 28 wt % sludge oxides, with the balance consisting of glass-forming chemicals, primarily glass frit. Borosilicate glass was chosen because of its stability toward reaction with potential repository groundwaters, its relatively high ability to incorporate nuclides found in the sludge into the solid matrix, and its reasonably low melting temperature. The glass frit contains approximately 71% SiO/sub 2/, 12% B/sub 2/O/sub 3/, and 10% Na/sub 2/O. Tests to quantify the stability of DWPF waste glass have been performed under a wide variety of conditions, including simulations of potential repository environments. Based on these tests, DWPF waste glass should easily meet repository criteria. The canister is filled with about 3700 lb of glass which occupies 85% of the free canister volume. The filled canister will generate approximately 690 watts when filled with oxides from 5-year-old sludge and precipitate from 15-year-old supernate. The radionuclide activity of the canister is about 233,000 curies, with an estimated radiation level of 5600 rad/hour at the canister surface. 14 figs., 28 tabs.

  15. Using Ionic Liquids in Selective Hydrocarbon Conversion Processes

    SciTech Connect (OSTI)

    Tang, Yongchun; Periana, Roy; Chen, Weiqun; van Duin, Adri; Nielsen, Robert; Shuler, Patrick; Ma, Qisheng; Blanco, Mario; Li, Zaiwei; Oxgaard, Jonas; Cheng, Jihong; Cheung, Sam; Pudar, Sanja

    2009-09-28T23:59:59.000Z

    This is the Final Report of the five-year project Using Ionic Liquids in Selective Hydrocarbon Conversion Processes (DE-FC36-04GO14276, July 1, 2004- June 30, 2009), in which we present our major accomplishments with detailed descriptions of our experimental and theoretical efforts. Upon the successful conduction of this project, we have followed our proposed breakdown work structure completing most of the technical tasks. Finally, we have developed and demonstrated several optimized homogenously catalytic methane conversion systems involving applications of novel ionic liquids, which present much more superior performance than the Catalytica system (the best-to-date system) in terms of three times higher reaction rates and longer catalysts lifetime and much stronger resistance to water deactivation. We have developed in-depth mechanistic understandings on the complicated chemistry involved in homogenously catalytic methane oxidation as well as developed the unique yet effective experimental protocols (reactors, analytical tools and screening methodologies) for achieving a highly efficient yet economically feasible and environmentally friendly catalytic methane conversion system. The most important findings have been published, patented as well as reported to DOE in this Final Report and our 20 Quarterly Reports.

  16. Composition and process for the encapsulation and stabilization of radioactive, hazardous and mixed wastes

    DOE Patents [OSTI]

    Kalb, P.D.; Colombo, P.

    1999-07-20T23:59:59.000Z

    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.

  17. Composition and process for the encapsulation and stabilization of radioactive, hazardous and mixed wastes

    DOE Patents [OSTI]

    Kalb, P.D.; Colombo, P.

    1998-03-24T23:59:59.000Z

    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.

  18. 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)

    1998-03-24T23:59:59.000Z

    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.

  19. Composition and process for the encapsulation and stabilization of radioactive hazardous and mixed wastes

    DOE Patents [OSTI]

    Kalb, P.D.; Colombo, P.

    1997-07-15T23:59:59.000Z

    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.

  20. 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)

    1997-01-01T23:59:59.000Z

    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.

  1. 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)

    1999-07-20T23:59:59.000Z

    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.

  2. Voluntary Protection Program Onsite Review, Transuranic Waste Processing Center- March 2008

    Broader source: Energy.gov [DOE]

    Evaluation to determine whether EnergX, LLC Transuranic Waste Processing Centeris continuing to perform at a level deserving DOE-VPP Star recognition.

  3. Voluntary Protection Program Onsite Review, Transuranic Waste Processing Center- May 2009

    Broader source: Energy.gov [DOE]

    Evaluation to determine whether Transuranic Waste Processing Center is continuing to perform at a level deserving DOE-VPP Star recognition.

  4. WASTE PROCESSING ANNUAL NUCLEAR SAFETY RELATED R AND D REPORT FOR CY2008

    SciTech Connect (OSTI)

    Fellinger, A.

    2009-10-15T23:59:59.000Z

    The Engineering and Technology Office of Waste Processing identifies and reduces engineering and technical risks associated with key waste processing project decisions. The risks, and actions taken to mitigate those risks, are determined through technology readiness assessments, program reviews, technology information exchanges, external technical reviews, technical assistance, and targeted technology development and deployment (TDD). The Office of Waste Processing TDD program prioritizes and approves research and development scopes of work that address nuclear safety related to processing of highly radioactive nuclear wastes. Thirteen of the thirty-five R&D approved work scopes in FY2009 relate directly to nuclear safety, and are presented in this report.

  5. An Istrument for Measuring the TRU Concentration in High-Level Liquid Waste

    SciTech Connect (OSTI)

    Brodzinski, Ronald L.; Craig, R. A.; Fink, Samuel D.; Hensley, Walter K.; Holt, Noah O.; Knopf, Michael A.; Lepel, Elwood A.; Mullen, O Dennis; Salaymeh, Saleem R.; Samuel, Todd J.; Smart, John E.; Tinker, Michael R.; Walker, Darrell D.

    2005-02-01T23:59:59.000Z

    An online monitor has been designed, built, and tested, which is capable of measuring the residual transuranic concentrations in processed high-level wastes with a detection limit of 370 Bq/ml (10 nCi/ml) in less than six hours. The monitor measures the neutrons produced by the transuranics, primarily via (?,n) reactions, in the presence of gamma-ray fields up to 1 Sv/h (100 R/h). The optimum design was determined by Monte Carlo modeling and then tempered with practical engineering and cost considerations. Correct operation of the monitor was demonstrated in a hot cell utilizing an actual sample of high-level waste. Results of that demonstration are given, and suggestions for improvements in the next generation system are discussed.

  6. Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source

    E-Print Network [OSTI]

    Belanger, David P.

    1 Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source renewed interest amongst the nuclear science community as the debate over nuclear waste has increased .................................................................................27 2.1.2 Waste Minimization

  7. HWMA/RCRA Closure Plan for the CPP-648 Radioactive Solid and Liquid Waste Storage Tank System (VES-SFE-106)

    SciTech Connect (OSTI)

    S. K. Evans

    2006-08-15T23:59:59.000Z

    This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan for the Radioactive Solid and Liquid Waste Storage Tank System located in the adjacent to the Sludge Tank Control House (CPP-648), Idaho Nuclear Technology and Engineering Center, Idaho National Laboratory, was developed to meet the interim status closure requirements for a tank system. The system to be closed includes a tank and associated ancillary equipment that were determined to have managed hazardous waste. The CPP-648 Radioactive Solid and Liquid Waste Storage Tank System will be "cleaned closed" in accordance with the requirements of the Hazardous Waste Management Act/Resource Conservation and Recovery Act as implemented by the Idaho Administrative Procedures Act and 40 Code of Federal Regulations 265. This closure plan presents the closure performance standards and methods of acheiving those standards for the CPP-648 Radioactive Solid and Liquid Waste Storage Tank System.

  8. Hanford Site Secondary Waste Roadmap

    SciTech Connect (OSTI)

    Westsik, Joseph H.

    2009-01-29T23:59:59.000Z

    Summary The U.S. Department of Energy (DOE) is making plans to dispose of 54 million gallons of radioactive tank wastes at the Hanford Site near Richland, Washington. The high-level wastes and low-activity wastes will be vitrified and placed in permanent disposal sites. Processing of the tank wastes will generate secondary wastes, including routine solid wastes and liquid process effluents, and these need to be processed and disposed of also. The Department of Energy Office of Waste Processing sponsored a meeting to develop a roadmap to outline the steps necessary to design the secondary waste forms. Representatives from DOE, the U.S. Environmental Protection Agency, the Washington State Department of Ecology, the Oregon Department of Energy, Nuclear Regulatory Commission, technical experts from the DOE national laboratories, academia, and private consultants convened in Richland, Washington, during the week of July 21-23, 2008, to participate in a workshop to identify the risks and uncertainties associated with the treatment and disposal of the secondary wastes and to develop a roadmap for addressing those risks and uncertainties. This report describes the results of the roadmap meeting in Richland. Processing of the tank wastes will generate secondary wastes, including routine solid wastes and liquid process effluents. The secondary waste roadmap workshop focused on the waste streams that contained the largest fractions of the 129I and 99Tc that the Integrated Disposal Facility risk assessment analyses were showing to have the largest contribution to the estimated IDF disposal impacts to groundwater. Thus, the roadmapping effort was to focus on the scrubber/off-gas treatment liquids with 99Tc to be sent to the Effluent Treatment Facility for treatment and solidification and the silver mordenite and carbon beds with the captured 129I to be packaged and sent to the IDF. At the highest level, the secondary waste roadmap includes elements addressing regulatory and performance requirements, waste composition, preliminary waste form screening, waste form development, process design and support, and validation. The regulatory and performance requirements activity will provide the secondary waste-form performance requirements. The waste-composition activity will provide workable ranges of secondary waste compositions and formulations for simulants and surrogates. Preliminary waste form screening will identify candidate waste forms for immobilizing the secondary wastes. The waste form development activity will mature the waste forms, leading to a selected waste form(s) with a defensible understanding of the long-term release rate and input into the critical decision process for a secondary waste treatment process/facility. The process and design support activity will provide a reliable process flowsheet and input to support a robust facility design. The validation effort will confirm that the selected waste form meets regulatory requirements. The final outcome of the implementation of the secondary waste roadmap is the compliant, effective, timely, and cost-effective disposal of the secondary wastes. The work necessary to address the programmatic, regulatory, and technical risks and uncertainties identified through the Secondary Waste Roadmap Workshop are assembled into several program needs elements. Programmatic/Regulatory needs include: • Select and deploy Hanford tank waste supplemental treatment technology • Provide treatment capability for secondary waste streams from tank waste treatment • Develop consensus on secondary waste form acceptance. Technology needs include: • Define secondary waste composition ranges and uncertainties • Identify and develop waste forms for secondary waste immobilization and disposal • Develop test methods to characterize secondary waste form performance. Details for each of these program elements are provided.

  9. FRIT OPTIMIZATION FOR SLUDGE BATCH PROCESSING AT THE DEFENSE WASTE PROCESSING FACILITY

    SciTech Connect (OSTI)

    Fox, K.

    2009-01-28T23:59:59.000Z

    The Savannah River National Laboratory (SRNL) Frit Development Team recommends that the Defense Waste Processing Facility (DWPF) utilize Frit 418 for initial processing of high level waste (HLW) Sludge Batch 5 (SB5). The extended SB5 preparation time and need for DWPF feed have necessitated the use of a frit that is already included on the DWPF procurement specification. Frit 418 has been used previously in vitrification of Sludge Batches 3 and 4. Paper study assessments predict that Frit 418 will form an acceptable glass when combined with SB5 over a range of waste loadings (WLs), typically 30-41% based on nominal projected SB5 compositions. Frit 418 has a relatively high degree of robustness with regard to variation in the projected SB5 composition, particularly when the Na{sub 2}O concentration is varied. The acceptability (chemical durability) and model applicability of the Frit 418-SB5 system will be verified experimentally through a variability study, to be documented separately. Frit 418 has not been designed to provide an optimal melt rate with SB5, but is recommended for initial processing of SB5 until experimental testing to optimize a frit composition for melt rate can be completed. Melt rate performance can not be predicted at this time and must be determined experimentally. Note that melt rate testing may either identify an improved frit for SB5 processing (one which produces an acceptable glass at a faster rate than Frit 418) or confirm that Frit 418 is the best option.

  10. RECOMMENDED FRIT COMPOSITION FOR INITIAL SLUDGE BATCH 5 PROCESSING AT THE DEFENSE WASTE PROCESSING FACILITY

    SciTech Connect (OSTI)

    Fox, K; Tommy Edwards, T; David Peeler, D

    2008-06-25T23:59:59.000Z

    The Savannah River National Laboratory (SRNL) Frit Development Team recommends that the Defense Waste Processing Facility (DWPF) utilize Frit 418 for initial processing of high level waste (HLW) Sludge Batch 5 (SB5). The extended SB5 preparation time and need for DWPF feed have necessitated the use of a frit that is already included on the DWPF procurement specification. Frit 418 has been used previously in vitrification of Sludge Batches 3 and 4. Paper study assessments predict that Frit 418 will form an acceptable glass when combined with SB5 over a range of waste loadings (WLs), typically 30-41% based on nominal projected SB5 compositions. Frit 418 has a relatively high degree of robustness with regard to variation in the projected SB5 composition, particularly when the Na{sub 2}O concentration is varied. The acceptability (chemical durability) and model applicability of the Frit 418-SB5 system will be verified experimentally through a variability study, to be documented separately. Frit 418 has not been designed to provide an optimal melt rate with SB5, but is recommended for initial processing of SB5 until experimental testing to optimize a frit composition for melt rate can be completed. Melt rate performance can not be predicted at this time and must be determined experimentally. Note that melt rate testing may either identify an improved frit for SB5 processing (one which produces an acceptable glass at a faster rate than Frit 418) or confirm that Frit 418 is the best option.

  11. Waste receiving and processing plant control system; system design description

    SciTech Connect (OSTI)

    LANE, M.P.

    1999-02-24T23:59:59.000Z

    The Plant Control System (PCS) is a heterogeneous computer system composed of numerous sub-systems. The PCS represents every major computer system that is used to support operation of the Waste Receiving and Processing (WRAP) facility. This document, the System Design Description (PCS SDD), includes several chapters and appendices. Each chapter is devoted to a separate PCS sub-system. Typically, each chapter includes an overview description of the system, a list of associated documents related to operation of that system, and a detailed description of relevant system features. Each appendice provides configuration information for selected PCS sub-systems. The appendices are designed as separate sections to assist in maintaining this document due to frequent changes in system configurations. This document is intended to serve as the primary reference for configuration of PCS computer systems. The use of this document is further described in the WRAP System Configuration Management Plan, WMH-350, Section 4.1.

  12. RESULTS OF THE EXTRACTION-SCRUB-STRIP TESTING USING AN IMPROVED SOLVENT FORMULATION AND SALT WASTE PROCESSING FACILITY SIMULATED WASTE

    SciTech Connect (OSTI)

    Peters, T.; Washington, A.; Fink, S.

    2012-01-09T23:59:59.000Z

    The Office of Waste Processing, within the Office of Technology Innovation and Development, is funding the development of an enhanced solvent - also known as the next generation solvent (NGS) - for deployment at the Savannah River Site to remove cesium from High Level Waste. The technical effort is a collaborative effort between Oak Ridge National Laboratory (ORNL) and Savannah River National Laboratory (SRNL). As part of the program, the Savannah River National Laboratory (SRNL) has performed a number of Extraction-Scrub-Strip (ESS) tests. These batch contact tests serve as first indicators of the cesium mass transfer solvent performance with actual or simulated waste. The test detailed in this report used simulated Tank 49H material, with the addition of extra potassium. The potassium was added at 1677 mg/L, the maximum projected (i.e., a worst case feed scenario) value for the Salt Waste Processing Facility (SWPF). The results of the test gave favorable results given that the potassium concentration was elevated (1677 mg/L compared to the current 513 mg/L). The cesium distribution value, DCs, for extraction was 57.1. As a comparison, a typical D{sub Cs} in an ESS test, using the baseline solvent formulation and the typical waste feed, is {approx}15. The Modular Caustic Side Solvent Extraction Unit (MCU) uses the Caustic-Side Solvent Extraction (CSSX) process to remove cesium (Cs) from alkaline waste. This process involves the use of an organic extractant, BoBCalixC6, in an organic matrix to selectively remove cesium from the caustic waste. The organic solvent mixture flows counter-current to the caustic aqueous waste stream within centrifugal contactors. After extracting the cesium, the loaded solvent is stripped of cesium by contact with dilute nitric acid and the cesium concentrate is transferred to the Defense Waste Processing Facility (DWPF), while the organic solvent is cleaned and recycled for further use. The Salt Waste Processing Facility (SWPF), under construction, will use the same process chemistry. The Office of Waste Processing (EM-31) expressed an interest in investigating the further optimization of the organic solvent by replacing the BoBCalixC6 extractant with a more efficient extractant. This replacement should yield dividends in improving cesium removal from the caustic waste stream, and in the rate at which the caustic waste can be processed. To that end, EM-31 provided funding for both the Savannah River National Laboratory (SRNL) and the Oak Ridge National Laboratory (ORNL). SRNL wrote a Task Technical Quality and Assurance Plan for this work. As part of the envisioned testing regime, it was decided to perform an ESS test using a simulated waste that simulated a typical envisioned SWPF feed, but with added potassium to make the waste more challenging. Potassium interferes in the cesium removal, and its concentration is limited in the feed to <1950 mg/L. The feed to MCU has typically contained <500 mg/L of potassium.

  13. Investigation of redox processes at semiconductor electrode liquid junctions

    SciTech Connect (OSTI)

    Koval, C.A.

    1990-08-01T23:59:59.000Z

    Research in fundamental aspects of photoelectrochemical cells has been in the following areas: chemical probes for hot carrier processes, electrostatic theory for describing electrical interactions at interfaces, and kinetics of electron transfer at ideal semiconductor solution interfaces. Our goal is to achieve a better understanding of dark and photo-induced current flow at the semiconductor electrode/redox electrolyte interface (SEI) so that devices and processes utilizing this interface for solar energy conversion can be developed or improved. Our most important accomplishment has been the development of a redox system capable of detecting hot electrons at the p-InP/acetonitrile interface. Also, we have examined electrostatic theory for the image potential of an ion as a function of distance from the SEI. Finally, our group was one of the first to realize that the 2-dimensional metal chalcogenides (MC) are excellent materials for fundamental studies of electron transfer at the SEI. One of the chief potential advantages for use of MC's is the formation of semiconductor/liquid junctions with nearly ideal electrochemical properties. 27 refs., 1 fig.

  14. The PERC{trademark} process: Existing and potential applications for induction coupled plasma technology in hazardous and radioactive waste treatment

    SciTech Connect (OSTI)

    Blutke, A.S.; Vavruska, J.S.; Serino, J.F. [Plasma Technology, Inc., Santa Fe, NM (United States)

    1996-12-31T23:59:59.000Z

    Plasma Technology, Inc. (PTI), a Santa Fe, New Mexico corporation has developed the Plasma Energy Recycle and Conversion (PERC){trademark} treatment process as a safe and environmentally clean alternative to conventional thermal destruction technologies. The PERC{trademark} treatment process uses as its heat source an advanced Induction Coupled Plasma (ICP) torch connected to a reaction chamber system with an additional emission control system. For example, organic-based gas, liquid, slurry, and/or solid waste streams can be converted into usable or even salable products while residual emissions are reduced to an absolute minimum. In applications for treatment of hazardous and radioactive waste streams, the PERC system could be used for destruction of the hazardous organic constituents and/or significant waste volume reduction while capturing the radioactive fraction in a non-leachable form. Like Direct Current (DC) and Alternating Current (AC) arc plasma systems, ICP torches offer sufficient energy to decompose, melt and/or vitrify any waste stream. The decision for an arc plasma or an IC plasma system has to be made on a case by case evaluation and is highly dependent on the specific waste stream`s form and composition. Induction coupled plasma technology offers one simple, but significant difference compared to DC or AC arc plasma systems: the ICP torch is electrodeless. To date, enormous research effort has been spent to improve the lifetime of electrodes and the effectiveness of related cooling systems. Arc plasma systems are established in research laboratories worldwide and are approaching a broad use in commercial applications. ICP technology has been improved relatively recently, but nowadays offers complete new and beneficial approaches in the field of waste conversion and treatment.

  15. 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.

    2011-09-12T23:59:59.000Z

    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.

  16. Cold End Inserts for Process Gas Waste Heat Boilers Air Products, operates hydrogen production plants, which utilize large waste heat boilers (WHB)

    E-Print Network [OSTI]

    Demirel, Melik C.

    Cold End Inserts for Process Gas Waste Heat Boilers Overview Air Products, operates hydrogen production plants, which utilize large waste heat boilers (WHB) to cool process syngas. The gas enters satisfies all 3 design criteria. · Correlations relating our experimental results to a waste heat boiler

  17. Process Description for the Retrieval of Earth Covered Transuranic (TRU) Waste Containers at the Hanford Site

    SciTech Connect (OSTI)

    DEROSA, D.C.

    2000-01-13T23:59:59.000Z

    This document describes process and operational options for retrieval of the contact-handled suspect transuranic waste drums currently stored below grade in earth-covered trenches at the Hanford Site. Retrieval processes and options discussed include excavation, container retrieval, venting, non-destructive assay, criticality avoidance, incidental waste handling, site preparation, equipment, and shipping.

  18. Thermodynamic estimation of minor element distribution between immiscible liquids in Fe-Cu-based metal phase generated in melting treatment of municipal solid wastes

    SciTech Connect (OSTI)

    Lu, X. [School of Metallurgical and Ecological Engineering, The University of Science and Technology Beijing, 30 Xueyuan Road, Haidian District, Beijing 100083 (China); Nakajima, K.; Sakanakura, H. [Research Center for Material Cycles and Waste Management, National Institute for Environmental Studies (NIES), 16-2 Onogawa, Tsukuba 305-8506 (Japan); Matsubae, K. [Graduate School of Engineering, Tohoku University, 6-6-11 Aza-Aoba, Aramaki, Sendai 980-8579 (Japan); Bai, H. [School of Metallurgical and Ecological Engineering, The University of Science and Technology Beijing, 30 Xueyuan Road, Haidian District, Beijing 100083 (China); Nagasaka, T., E-mail: t-nagasaka@m.tohoku.ac.jp [Graduate School of Engineering, Tohoku University, 6-6-11 Aza-Aoba, Aramaki, Sendai 980-8579 (Japan)

    2012-06-15T23:59:59.000Z

    Graphical abstract: Display Omitted Highlights: Black-Right-Pointing-Pointer Two liquids separation of metal occurs in the melting of municipal solid waste. Black-Right-Pointing-Pointer The distribution of PGMs etc. between two liquid metal phases is studied. Black-Right-Pointing-Pointer Quite simple thermodynamic model is applied to predict the distribution ratio. Black-Right-Pointing-Pointer Au and Ag originated from WEEE are found to be concentrated into Cu-rich phase. - Abstract: Waste electrical and electronic equipment (WEEE) has become an important target in managing material cycles from the viewpoint of not only waste management and control of environmental pollution but also resource conservation. This study investigated the distribution tendency of trace elements in municipal solid waste (MSW) or incinerator ash, including valuable non-ferrous metals (Ni, Co, Cr, Mn, Mo, Ti, V, W, Zr), precious group metals (PGMs) originated from WEEE (Ag, Au, Pd, Pt), and others (Al, B, Pb, Si), between Fe-rich and Cu-rich metal phases by means of simple thermodynamic calculations. Most of the typical alloying elements for steel (Co, Cr, Mo, Nb, Ni, Si, Ti, V, and W) and Rh were preferentially distributed into the Fe-rich phase. PGMs, such as Au, Ag, and Pd, were enriched in the Cu-rich phase, whereas Pt was almost equally distributed into both phases. Since the primary metallurgical processing of Cu is followed by an electrolysis for refining, and since PGMs in crude copper have been industrially recovered from the resulting anode slime, our results indicated that Ag, Au, and Pd could be effectively recovered from MSW if the Cu-rich phase could be selectively collected.

  19. Physical Properties Models for Simulation of Processes to Treat INEEL Tank Farm Waste: Thermodynamic Equilibrium

    SciTech Connect (OSTI)

    Nichols, Todd Travis; Taylor, Dean Dalton

    2002-07-01T23:59:59.000Z

    A status is presented of the development during FY2002 of a database for physical properties models for the simulation of the treatment of Sodium-Bearing Waste (SBW) at the Idaho National Engineering and Environmental Laboratory. An activity coefficient model is needed for concentrated, aqueous, multi-electrolyte solutions that can be used by process design practitioners. Reasonable first-order estimates of activity coefficients in the relevant media are needed rather than an incremental improvement in theoretical approaches which are not usable by practitioners. A comparison of the Electrolyte Non-Random Two-Liquid (ENRTL) and Pitzer ion-interaction models for the thermodynamic representation of SBW is presented. It is concluded that Pitzer's model is superior to ENRTL in modeling treatment processes for SBW. The applicability of the Pitzer treatment to high concentrations of pertinent species and to the determination of solubilities and chemical equilibria is addressed. Alternate values of Pitzer parameters for HCl, H2SO4, and HNO3 are proposed, applicable up to 16m, and 12m, respectively. Partial validation of the implementation of Pitzer's treatment within the commercial process simulator ASPEN Plus was performed.

  20. Physical Properties Models for Simulation of Processes to Treat INEEL Tank Farm Waste: Thermodynamic Equilibrium

    SciTech Connect (OSTI)

    Nichols, T.T.; Taylor, D.D.

    2002-07-18T23:59:59.000Z

    A status is presented of the development during FY2002 of a database for physical properties models for the simulation of the treatment of Sodium-Bearing Waste (SBW) at the Idaho National Engineering and Environmental Laboratory. An activity coefficient model is needed for concentrated, aqueous, multi-electrolyte solutions that can be used by process design practitioners. Reasonable first-order estimates of activity coefficients in the relevant media are needed rather than an incremental improvement in theoretical approaches which are not usable by practitioners. A comparison of the Electrolyte Non-Random Two-Liquid (ENRTL) and Pitzer ion-interaction models for the thermodynamic representation of SBW is presented. It is concluded that Pitzer's model is superior to ENRTL in modeling treatment processes for SBW. The applicability of the Pitzer treatment to high concentrations of pertinent species and to the determination of solubilities and chemical equilibria is addressed. Alternate values of Pitzer parameters for HCl, H2SO4, and HNO3 are proposed, applicable up to 16m, and 12m, respectively. Partial validation of the implementation of Pitzer's treatment within the commercial process simulator ASPEN Plus was performed.

  1. Processing and waste disposal representative for fusion breeder blanket systems

    SciTech Connect (OSTI)

    Finn, P.A.; Vogler, S.

    1987-01-01T23:59:59.000Z

    This study is an evaluation of the waste handling concepts applicable to fusion breeder systems. Its goal is to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under US Nuclear Regulatory regulations. The radionuclides expected in the materials used in fusion reactor blankets are described, as are plans for reprocessing and disposal of the components of different breeder blankets. An estimate of the operating costs involved in waste disposal is made.

  2. Using Waste Heat for External Processes (English/Chinese) (Fact Sheet)

    SciTech Connect (OSTI)

    Not Available

    2011-10-01T23:59:59.000Z

    Chinese translation of the Using Waste Heat for External Processes fact sheet. Provides suggestions on how to use waste heat in industrial applications. The temperature of exhaust gases from fuel-fired industrial processes depends mainly on the process temperature and the waste heat recovery method. Figure 1 shows the heat lost in exhaust gases at various exhaust gas temperatures and percentages of excess air. Energy from gases exhausted from higher temperature processes (primary processes) can be recovered and used for lower temperature processes (secondary processes). One example is to generate steam using waste heat boilers for the fluid heaters used in petroleum crude processing. In addition, many companies install heat exchangers on the exhaust stacks of furnaces and ovens to produce hot water or to generate hot air for space heating.

  3. agriculture process waste: Topics by E-print Network

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

    (more) Radhika M 2014-01-01 117 Waste to Energy (WTE): Conventional and Plasma-assisted Gasification - Experimental and Modeling Studies. Open Access Theses and Dissertations...

  4. Isolation of Metals from Liquid Wastes: Reactive Scavenging in Turbulent Thermal Reactors

    SciTech Connect (OSTI)

    William Linak

    2004-12-16T23:59:59.000Z

    Sorption of cesium and strontium on kaolinite powders was investigated as a means to minimize the emissions of these metals during certain high temperature processes currently being developed to isolate and dispose of radiological and mixed wastes. In this work, non-radioactive aqueous cesium acetate or strontium acetate was atomized down the center of a natural gas flame supported on a variable-swirl burner in a refractory-lined laboratory-scale combustion facility. Kaolinite powder was injected at a post-flame location in the combustor. Cesium readily vaporizes in the high temperature regions of the combustor, but was reactively scavenged onto dispersed kaolinite. Global sorption mechanisms of cesium vapor on kaolinite were quantified, and are related to those available in the literature for sodium and lead. Both metal adsorption and substrate deactivation steps are important, and so there is an optimum temperature, between 1400 and 1500 K, at which maximum sorption occurs. The presence of chlorine inhibits cesium sorption. In contrast to cesium, and in the absence of chlorine, strontium was only partially vaporized and was, therefore, only partially scavengeable. The strontium data did not allow quantification of global kinetic mechanisms of interaction, although equilibrium arguments provided insight into the effects of chlorine on strontium sorption. These results have implications for the use of sorbents to control cesium and strontium emissions during high temperature waste processing including incineration and vitrification.

  5. Liquid Resin Infusion process monitoring with superimposed Fibre Bragg Grating sensor

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 Liquid Resin Infusion process monitoring with superimposed Fibre Bragg Grating sensor Emmanuel Resin Infusion (LRI) , with the FBG/LPG sensor embedded in a composite part. Dielectric analysis the material and the structure. Among the various composite manufacturing processes, Liquid Resin Infusion (LRI

  6. Selection of Controlled Variables for a Natural Gas to Liquids Process Mehdi Panahi and Sigurd Skogestad*

    E-Print Network [OSTI]

    Skogestad, Sigurd

    Selection of Controlled Variables for a Natural Gas to Liquids Process Mehdi Panahi and Sigurd variables (CVs) for a natural gas to hydrocarbon liquids (GTL) process based on the idea of self of operation are studied. In mode I, where the natural gas flow rate is given, there are three unconstrained

  7. Hanford Waste Vitrification Plant full-scale feed preparation testing with water and process simulant slurries

    SciTech Connect (OSTI)

    Gaskill, J.R.; Larson, D.E.; Abrigo, G.P. [and others] [and others

    1996-03-01T23:59:59.000Z

    The Hanford Waste Vitrification Plant was intended to convert selected, pretreated defense high-level waste and transuranic waste from the Hanford Site into a borosilicate glass. A full-scale testing program was conducted with nonradioactive waste simulants to develop information for process and equipment design of the feed-preparation system. The equipment systems tested included the Slurry Receipt and Adjustment Tank, Slurry Mix Evaporator, and Melter-Feed Tank. The areas of data generation included heat transfer (boiling, heating, and cooling), slurry mixing, slurry pumping and transport, slurry sampling, and process chemistry. 13 refs., 129 figs., 68 tabs.

  8. Comparison between CMPO and DHDECMP for alpha decontamination of radioactive liquid waste

    SciTech Connect (OSTI)

    Muscatello, A.C.; Yarbro, S.L.; Marsh, S.F.

    1990-01-01T23:59:59.000Z

    Ion exchange is the major method used at Los Alamos to recover and purify plutonium from a variety of different contaminants. During this process, a high-acid (5-7M), low-activity stream is produced that presently is concentrated by evaporation, then cemented for long-term disposal. Our goal is to remove and concentrate the radioactive elements so that the remainder can be treated as low-level'' or regular industrial waste. Solvent extraction with neutral bifunctional extractants, such as DHDECMP and CMPO, has been chosen as the process to be developed. Experimental work has shown that both extractants effectively remove actinides to below the required limits, but that CMPO was much more difficult to strip. In addition, studies of plutonium and americium removal using a wide variety of ion exchangers and supported extractants including DHDECMP, CMPO, and TOPO will be reviewed. 22 refs., 10 figs., 3 tabs.

  9. STATISTICAL SAMPLING FOR IN-SERVICE INSPECTION OF LIQUID WASTE TANKS AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Harris, S.; Baxter, L.

    2011-04-07T23:59:59.000Z

    Savannah River Remediation, LLC (SRR) is implementing a statistical sampling strategy for In-Service Inspection (ISI) of Liquid Waste (LW) Tanks at the United States Department of Energy's Savannah River Site (SRS) in Aiken, South Carolina. As a component of SRS's corrosion control program, the ISI program assesses tank wall structural integrity through the use of ultrasonic testing (UT). The statistical strategy for ISI is based on the random sampling of a number of vertically oriented unit areas, called strips, within each tank. The number of strips to inspect was determined so as to attain, over time, a high probability of observing at least one of the worst 5% in terms of pitting and corrosion across all tanks. The probability estimation to determine the number of strips to inspect was performed using the hypergeometric distribution. Statistical tolerance limits for pit depth and corrosion rates were calculated by fitting the lognormal distribution to the data. In addition to the strip sampling strategy, a single strip within each tank was identified to serve as the baseline for a longitudinal assessment of the tank safe operational life. The statistical sampling strategy enables the ISI program to develop individual profiles of LW tank wall structural integrity that collectively provide a high confidence in their safety and integrity over operational lifetimes.

  10. Los Alamos Waste Management Cost Estimation Model; Final report: Documentation of waste management process, development of Cost Estimation Model, and model reference manual

    SciTech Connect (OSTI)

    Matysiak, L.M.; Burns, M.L.

    1994-03-01T23:59:59.000Z

    This final report completes the Los Alamos Waste Management Cost Estimation Project, and includes the documentation of the waste management processes at Los Alamos National Laboratory (LANL) for hazardous, mixed, low-level radioactive solid and transuranic waste, development of the cost estimation model and a user reference manual. The ultimate goal of this effort was to develop an estimate of the life cycle costs for the aforementioned waste types. The Cost Estimation Model is a tool that can be used to calculate the costs of waste management at LANL for the aforementioned waste types, under several different scenarios. Each waste category at LANL is managed in a separate fashion, according to Department of Energy requirements and state and federal regulations. The cost of the waste management process for each waste category has not previously been well documented. In particular, the costs associated with the handling, treatment and storage of the waste have not been well understood. It is anticipated that greater knowledge of these costs will encourage waste generators at the Laboratory to apply waste minimization techniques to current operations. Expected benefits of waste minimization are a reduction in waste volume, decrease in liability and lower waste management costs.

  11. Process and material that encapsulates solid hazardous waste

    DOE Patents [OSTI]

    O'Brien, Michael H. (Idaho Falls, ID); Erickson, Arnold W. (Idaho Falls, ID)

    1999-01-01T23:59:59.000Z

    A method of encapsulating mixed waste in which a thermoplastic polymer having a melting temperature less than about 150.degree. C. and sulfur and mixed waste are mixed at an elevated temperature not greater than about 200.degree. C. and mixed for a time sufficient to intimately mix the constituents, and then cooled to a solid. The resulting solid is also disclosed.

  12. Accepting Mixed Waste as Alternate Feed Material for Processing and Disposal at a Licensed Uranium Mill

    SciTech Connect (OSTI)

    Frydenland, D. C.; Hochstein, R. F.; Thompson, A. J.

    2002-02-26T23:59:59.000Z

    Certain categories of mixed wastes that contain recoverable amounts of natural uranium can be processed for the recovery of valuable uranium, alone or together with other metals, at licensed uranium mills, and the resulting tailings permanently disposed of as 11e.(2) byproduct material in the mill's tailings impoundment, as an alternative to treatment and/or direct disposal at a mixed waste disposal facility. This paper discusses the regulatory background applicable to hazardous wastes, mixed wastes and uranium mills and, in particular, NRC's Alternate Feed Guidance under which alternate feed materials that contain certain types of mixed wastes may be processed and disposed of at uranium mills. The paper discusses the way in which the Alternate Feed Guidance has been interpreted in the past with respect to processing mixed wastes and the significance of recent changes in NRC's interpretation of the Alternate Feed Guidance that sets the stage for a broader range of mixed waste materials to be processed as alternate feed materials. The paper also reviews the le gal rationale and policy reasons why materials that would otherwise have to be treated and/or disposed of as mixed waste, at a mixed waste disposal facility, are exempt from RCRA when reprocessed as alternate feed material at a uranium mill and become subject to the sole jurisdiction of NRC, and some of the reasons why processing mixed wastes as alternate feed materials at uranium mills is preferable to direct disposal. Finally, the paper concludes with a discussion of the specific acceptance, characterization and certification requirements applicable to alternate feed materials and mixed wastes at International Uranium (USA) Corporation's White Mesa Mill, which has been the most active uranium mill in the processing of alternate feed materials under the Alternate Feed Guidance.

  13. EIS-0081: Long-Term Management of Liquid High-Level Radioactive Waste Stored at Western New York Nuclear Service Center, West Valley, New York

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy Office of Terminal Waste Disposal and Remedial Action prepared this environmental impact statement to analyze the environmental and socioeconomic impacts resulting from the Department’s proposed action to construct and operate facilities necessary to solidify the liquid high-level wastes currently stored in underground tanks at West Valley, New York.

  14. 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

    2003-02-25T23:59:59.000Z

    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.

  15. Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

    Michael F. Simpson; Michael N. Patterson; Joon Lee; Yifeng Wang; Joshua Versey; Ammon Williams; Supathorn Phongikaroon; James Allensworth; Man-Sung Yim

    2013-10-01T23:59:59.000Z

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

  16. Management of salt waste from electrochemical processing of used nuclear fuel

    SciTech Connect (OSTI)

    Simpson, M.F.; Patterson, M.N. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415 (United States); Lee, J.; Wang, Y. [Sandia National Laboratory, Albuquerque, NM (United States); Versey, J.; Phongikaroon, S. [University of Idaho, Idaho Falls, ID (United States)

    2013-07-01T23:59:59.000Z

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electro-refiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form. (authors)

  17. Transfer Lines to Connect Liquid Waste Facilities and Salt Waste Processing

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn April 23,EnergyChicopeeTechnologyfact sheet

  18. Membrane contactor assisted extraction/reaction process employing ionic liquids

    DOE Patents [OSTI]

    Lin, Yupo J. (Naperville, IL); Snyder, Seth W. (Lincolnwood, IL)

    2012-02-07T23:59:59.000Z

    The present invention relates to a functionalized membrane contactor extraction/reaction system and method for extracting target species from multi-phase solutions utilizing ionic liquids. One preferred embodiment of the invented method and system relates to an extraction/reaction system wherein the ionic liquid extraction solutions act as both extraction solutions and reaction mediums, and allow simultaneous separation/reactions not possible with prior art technology.

  19. 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)

    2008-07-01T23:59:59.000Z

    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)

  20. An Instrument for Measuring the TRU Concentration in High-Level Liquid Waste

    SciTech Connect (OSTI)

    Brodzinski, Ronald L.; Craig, R A.; Fink, Samuel D.; Hensley, Walter K.; Holt, Noah OA; Knopf, Michael A.; Lepel, Elwood A.; Mullen, O Dennis; Salaymeh, Saleem R.; Samuel, Todd J.; Smart, John E.; Tinker, Mike R.; Walker, D

    2005-02-01T23:59:59.000Z

    An online monitor has been designed, built, and tested that is capable of measuring the residual transuranic concentrations in processed high-level wastes with a detection limit of 370 Bq/ml (10 nCi/ml) in less than six hours. The monitor measures the ({alpha},n) neutrons in the presence of gamma-ray fields up to 1 Sv/h (100 R/h). The optimum design was determined by Monte Carlo modeling and then tempered with practical engineering and cost considerations. A multiplicity counter is used in data acquisition to reject the large fraction of coincident and highly variable cosmic-ray-engendered background events and results in a S/N ratio {approx}1.

  1. Report on the flowsheet model for the electrochemical treatment of liquid radioactive wastes

    SciTech Connect (OSTI)

    Hobbs, D.T.

    1995-04-11T23:59:59.000Z

    The objective of this report is to describe the modeling and optimization procedure for the electrochemical removal of nitrates and nitrites from low level radioactive wastes. The simulation is carried out in SPEEDUP{trademark}, which is a state of the art flowsheet modeling package. The flowsheet model will provide a better understanding of the process and aid in the scale-up of the system. For example, the flowsheet model has shown that the electrochemical cell must be operated in batch mode to achieve 95% destruction. The present status of the flowsheet model is detailed in this report along with a systematic description of the batch optimization of the electrochemical cell. Results from two batch runs and one optimization run are also presented.

  2. Process for immobilizing plutonium into vitreous ceramic waste forms

    DOE Patents [OSTI]

    Feng, X.; Einziger, R.E.

    1997-08-12T23:59:59.000Z

    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.

  3. Process for immobilizing plutonium into vitreous ceramic waste forms

    DOE Patents [OSTI]

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

    1997-01-01T23:59:59.000Z

    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.

  4. Process for immobilizing plutonium into vitreous ceramic waste forms

    DOE Patents [OSTI]

    Feng, X.; Einziger, R.E.

    1997-01-28T23:59:59.000Z

    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.

  5. Identification of existing waste heat recovery and process improvement technologies

    SciTech Connect (OSTI)

    Watts, R.L.; Dodge, R.E.; Smith, S.A.; Ames, K.R.

    1984-03-01T23:59:59.000Z

    General information is provided on waste heat recovery opportunities. The currently available equipment for high- and low-temperature applications are described. Other equipment related to wasteheat recovery equipment such as components, instruments and controls, and cleaning equipment is discussed briefly. A description of the microcomputer data base is included. Suppliers of waste heat equipment are mentioned throughout the report, with specific contacts, addresses, and telephone numbers provided in an Appendix.

  6. Large Precipitate Hydrolysis Aqueous (PHA) Heel Process Development for the Defense Waste Processing Facility (DWPF)

    SciTech Connect (OSTI)

    Lambert, D.P. [Westinghouse Savannah River Company, AIKEN, SC (United States); Boley, C.S.; Jacobs, R.A.

    1998-06-04T23:59:59.000Z

    A modification to the Precipitate Hydrolysis flowsheet used in DWPF Waste Qualification Runs has been developed.

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

    E-Print Network [OSTI]

    Medvedev, Dmitry Gennadievich

    2005-11-01T23:59:59.000Z

    The processes of development of nuclear weapons resulted in accumulation of thousands of curies of high-level radioactive waste. Liquid waste produced in the US has been stored in carbon steel tanks in highly alkaline (1-3 ...

  8. UNITED STATES DEPARTMENT OF ENERGY OFFICE OF ENVIRONMENTAL MANAGEMENT WASTE PROCESSING ANNUAL TECHNOLOGY DEVELOPMENT REPORT 2008

    SciTech Connect (OSTI)

    Bush, S.

    2009-11-05T23:59:59.000Z

    The Office of Waste Processing identifies and reduces engineering and technical risks and uncertainties of the waste processing programs and projects of the Department of Energy's Environmental Management (EM) mission through the timely development of solutions to technical issues. The risks, and actions taken to mitigate those risks, are determined through technology readiness assessments, program reviews, technology information exchanges, external technical reviews, technical assistance, and targeted technology development and deployment. The Office of Waste Processing works with other DOE Headquarters offices and project and field organizations to proactively evaluate technical needs, identify multi-site solutions, and improve the technology and engineering associated with project and contract management. Participants in this program are empowered with the authority, resources, and training to implement their defined priorities, roles, and responsibilities. The Office of Waste Processing Multi-Year Program Plan (MYPP) supports the goals and objectives of the U.S. Department of Energy (DOE) - Office of Environmental Management Engineering and Technology Roadmap by providing direction for technology enhancement, development, and demonstration that will lead to a reduction of technical risks and uncertainties in EM waste processing activities. The MYPP summarizes the program areas and the scope of activities within each program area proposed for the next five years to improve safety and reduce costs and environmental impacts associated with waste processing; authorized budget levels will impact how much of the scope of activities can be executed, on a year-to-year basis. Waste Processing Program activities within the Roadmap and the MYPP are described in these seven program areas: (1) Improved Waste Storage Technology; (2) Reliable and Efficient Waste Retrieval Technologies; (3) Enhanced Tank Closure Processes; (4) Next-Generation Pretreatment Solutions; (5) Enhanced Stabilization Technologies; (6) Spent Nuclear Fuel; and (7) Challenging Materials. This report provides updates on 35 technology development tasks conducted during calendar year 2008 in the Roadmap and MYPP program areas.

  9. The release of technetium from defense waste processing facility glasses

    SciTech Connect (OSTI)

    Ebert, W.L.; Wolf, S.F.; Bates, J.K.

    1995-12-31T23:59:59.000Z

    Laboratory tests are being, conducted using two radionuclide-doped Defense Waste Processing, Facility (DWPF) glasses (referred to as SRL 13IA and SRL 202A) to characterize the effects of the glass surface area/solution volume (SN) ratio on the release and disposition of {Tc} and several actinide elements. Tests are being conducted at 90{degrees}C in a tuff ground water solution at SN ratios of 10, 2000, and 20,000 m{sup {minus}1} and have been completed through 1822 days. The formation of certain alteration phases in tests at 2000 and 20,000 m{sup {minus}1} results in an increase in the dissolution rates of both classes. The release of {Tc} parallels that of B and Na under most test conditions and its release increases when alteration phases form. However, in tests with SRL 202A glass at 20,000 ,{sup {minus}1}, the {Tc} concentration in solution decreases coincidentally with an increase in the nitrite/nitrate ratio that indicates a decrease in the solution Eh. This may have occurred due to radiolysis, glass dissolution, the formation of alteration phases, or vessel interactions. Technetium that was reduced from {Tc}(VII) to {Tc}(IV) may have precipitated, thou-h the amount of {Tc} was too low to detect any {Tc}-bearing phases. These results show the importance of conducting long-term tests with radioactive glasses to characterize the behavior of radionuclides, rather than relying on the observed behavior of nonradioactive surrogates.

  10. Hanford tank waste simulants specification and their applicability for the retrieval, pretreatment, and vitrification processes

    SciTech Connect (OSTI)

    GR Golcar; NG Colton; JG Darab; HD Smith

    2000-04-04T23:59:59.000Z

    A wide variety of waste simulants were developed over the past few years to test various retrieval, pretreatment and waste immobilization technologies and unit operations. Experiments can be performed cost-effectively using non-radioactive waste simulants in open laboratories. This document reviews the composition of many previously used waste simulants for remediation of tank wastes at the Hanford reservation. In this review, the simulants used in testing for the retrieval, pretreatment, and vitrification processes are compiled, and the representative chemical and physical characteristics of each simulant are specified. The retrieval and transport simulants may be useful for testing in-plant fluidic devices and in some cases for filtration technologies. The pretreatment simulants will be useful for filtration, Sr/TRU removal, and ion exchange testing. The vitrification simulants will be useful for testing melter, melter feed preparation technologies, and for waste form evaluations.

  11. Demonstrating Reliable High Level Waste Slurry Sampling Techniques to Support Hanford Waste Processing

    SciTech Connect (OSTI)

    Kelly, Steven E.

    2013-11-11T23:59:59.000Z

    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 capability using simulated Hanford High-Level Waste (HL W) formulations. This work represents one of the remaining technical issues with the high-level waste treatment mission at Hanford. The TOC must demonstrate the ability to adequately mix and sample high-level waste feed to meet the WTP Waste Acceptance Criteria and Data Quality Objectives. The sampling method employed must support both TOC and WTP requirements. To facilitate information transfer between the two facilities the mixing and sampling demonstrations are led by the One System Integrated Project Team. The One System team, Waste Feed Delivery Mixing and Sampling Program, has developed a full scale sampling loop to demonstrate sampler capability. This paper discusses the full scale sampling loops ability to meet precision and accuracy requirements, including lessons learned during testing. Results of the testing showed that the Isolok(R) sampler chosen for implementation provides precise, repeatable results. The Isolok(R) sampler accuracy as tested did not meet test success criteria. Review of test data and the test platform following testing by a sampling expert identified several issues regarding the sampler used to provide reference material used to judge the Isolok's accuracy. Recommendations were made to obtain new data to evaluate the sampler's accuracy utilizing a reference sampler that follows good sampling protocol.

  12. APPLICATION OF A THIN FILM EVAPORATOR SYSTEM FOR MANAGEMENT OF LIQUID HIGH-LEVEL WASTES AT HANFORD

    SciTech Connect (OSTI)

    TEDESCHI AR; WILSON RA

    2010-01-14T23:59:59.000Z

    A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORP/DOE), through Columbia Energy & Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper discusses results of pre-project pilot-scale testing by Columbia Energy and ongoing technology maturation development scope through fiscal year 2012, including planned additional pilot-scale and full-scale simulant testing and operation with actual radioactive tank waste.

  13. Adsorption/Membrane Filtration as a Contaminant Concentration and Separation Process for Mixed Wastes and Tank Wastes - Final Report

    SciTech Connect (OSTI)

    Benjamin, M.M.

    1999-10-01T23:59:59.000Z

    This project was conducted to evaluate novel approaches for removing radioactive strontium (Sr) and cesium (Cs) from the tank wastes. The bulk of the Sr removal research conducted as part of this project investigated adsorption of Sr onto a novel adsorbent known as iron-oxide-coated sand. The second major focus of the work was on the removal of cesium. Since the chemistries of strontium and cesium have little commonality, different materials (namely, cesium scavengers known as hexacyanoferrates, HCFs) were employed in these tests. This study bridged several scientific areas and yielded valuable knowledge for implementing new technological processes. The applicability of the results extends beyond the highly specialized application niches investigated experimentally to other issues of potential interest for EMSP programs (e.g., separation of chromium from a variety of wastes using IOCS, separation of Cs from neutral and acidic wastes with EC-controlled HCFs).

  14. Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario hydro`s Bruce nuclear generating station {open_quotes}A{close_quotes}

    SciTech Connect (OSTI)

    Day, J.E.; Baker, R.L. [ADTECHS Corporation, Herndon, VA (United States)

    1994-12-31T23:59:59.000Z

    Ontario Hydro at the Bruce Nuclear Generating Station {open_quotes}A{close_quotes} has undertaken a program to render the station`s liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology.

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

    SciTech Connect (OSTI)

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

    1999-11-11T23:59:59.000Z

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

  16. thermochemical conversion (TCC) process of swine manure as an alternative means of waste

    E-Print Network [OSTI]

    He, Brian

    substantially reduce the waste strength of swine manure slurry but also produce renewable energy. The end was historically linked to hydrogenation and other high-pressure thermal decomposition processes of coal

  17. Evaluation of Seafood Processing Wastes in Prepared Feeds for Red Drum (Sciaenops ocellatus)

    E-Print Network [OSTI]

    Pernu, Benjamin Mark

    2012-07-16T23:59:59.000Z

    High feed costs and increasing demand for fishmeal have intensified the search for alternative protein sources which are needed to allow world aquaculture to continue expanding. A severely underused marine resource is processing wastes of various...

  18. Economic comparison of centralizing or decentralizing processing facilities for defense transuranic waste

    SciTech Connect (OSTI)

    Brown, C M

    1980-07-01T23:59:59.000Z

    This study is part of a set of analyses under direction of the Transuranic Waste Management Program designed to provide comprehensive, systematic methodology and support necessary to better understand options for national long-term management of transuranic (TRU) waste. The report summarizes activities to evaluate the economics of possible alternatives in locating facilities to process DOE-managed transuranic waste. The options considered are: (1) Facilities located at all major DOE TRU waste generating sites. (2) Two or three regional facilities. (3) Central processing facility at only one DOE site. The study concludes that processing at only one facility is the lowest cost option, followed, in order of cost, by regional then individual site processing.

  19. Process Waste Heat Recovery in the Food Industry - A System Analysis

    E-Print Network [OSTI]

    Lundberg, W. L.; Mutone, G. A.

    1983-01-01T23:59:59.000Z

    An analysis of an industrial waste heat recovery system concept is discussed. For example purposes, a food processing plant operating an ammonia refrigeration system for storage and blast freezing is considered. Heat is withdrawn from...

  20. Radioactive Waste Conditioning, Immobilisation, And Encapsulation Processes And Technologies: Overview And Advances (Chapter 7)

    SciTech Connect (OSTI)

    Jantzen, Carol M. [Savannah River National Lab., Aiken SC (United States); Lee, William E. [Imperial College, London (United Kingdom). Dept. of Materials; Ojovan, Michael I. [Univ. of Sheffield (United Kingdom). Dept. of Materials Science and Engineering

    2012-10-19T23:59:59.000Z

    The main immobilization technologies that are available commercially and have been demonstrated to be viable are cementation, bituminization, and vitrification. Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either alkali borosilicate glass or alkali aluminophosphate glass. The exact compositions of nuclear waste glasses are tailored for easy preparation and melting, avoidance of glass-in-glass phase separation, avoidance of uncontrolled crystallization, and acceptable chemical durability, e.g., leach resistance. Glass has also been used to stabilize a variety of low level wastes (LLW) and mixed (radioactive and hazardous) low level wastes (MLLW) from other sources such as fuel rod cladding/decladding processes, chemical separations, radioactive sources, radioactive mill tailings, contaminated soils, medical research applications, and other commercial processes. The sources of radioactive waste generation are captured in other chapters in this book regarding the individual practices in various countries (legacy wastes, currently generated wastes, and future waste generation). Future waste generation is primarily driven by interest in sources of clean energy and this has led to an increased interest in advanced nuclear power production. The development of advanced wasteforms is a necessary component of the new nuclear power plant (NPP) flowsheets. Therefore, advanced nuclear wasteforms are being designed for robust disposal strategies. A brief summary is given of existing and advanced wasteforms: glass, glass-ceramics, glass composite materials (GCM’s), and crystalline ceramic (mineral) wasteforms that chemically incorporate radionuclides and hazardous species atomically in their structure. Cementitious, geopolymer, bitumen, and other encapsulant wasteforms and composites that atomically bond and encapsulate wastes are also discussed. The various processing technologies are cross-referenced to the various types of wasteforms since often a particular type of wasteform can be made by a variety of different processing technologies.

  1. Electromagnetic mixed waste processing system for asbestos decontamination

    SciTech Connect (OSTI)

    Kasevich, R.S.; Vaux, W.G.; Nocito, T.

    1995-12-01T23:59:59.000Z

    DOE sites contain a broad spectrum of asbestos materials (cloth, pipe lagging, sprayed insulation and other substances) which are contaminated with a combination of hazardous and radioactive wastes due to its use during the development of the U.S. nuclear weapons complex. These wastes consist of cutting oils, lubricants, solvents, PCB`s, heavy metals and radioactive contaminants. The radioactive contaminants are the activation, decay and fission products of DOE operations. The asbestos must be converted by removing and separating the hazardous and radioactive materials to prevent the formation of mixed wastes and to allow for both sanitary disposal and effective decontamination. Currently, no technology exists that can meet these sanitary and other objectives.

  2. Process for converting coal into liquid fuel and metallurgical coke

    DOE Patents [OSTI]

    Wolfe, Richard A. (Abingdon, VA); Im, Chang J. (Abingdon, VA); Wright, Robert E. (Bristol, TN)

    1994-01-01T23:59:59.000Z

    A method of recovering coal liquids and producing metallurgical coke utilizes low ash, low sulfur coal as a parent for a coal char formed by pyrolysis with a volatile content of less than 8%. The char is briquetted and heated in an inert gas over a prescribed heat history to yield a high strength briquette with less than 2% volatile content.

  3. Formulation and Characterization of Waste Glasses with Varying Processing Temperature

    SciTech Connect (OSTI)

    Kim, Dong-Sang; Schweiger, M. J.; Rodriguez, Carmen P.; Lepry, William C.; Lang, Jesse B.; Crum, Jarrod V.; Vienna, John D.; Johnson, Fabienne; Marra, James C.; Peeler, David K.

    2011-10-17T23:59:59.000Z

    This report documents the preliminary results of glass formulation and characterization accomplished within the finished scope of the EM-31 technology development tasks for WP-4 and WP-5, including WP-4.1.2: Glass Formulation for Next Generation Melter, WP-5.1.2.3: Systematic Glass Studies, and WP-5.1.2.4: Glass Formulation for Specific Wastes. This report also presents the suggested studies for eventual restart of these tasks. The initial glass formulation efforts for the cold crucible induction melter (CCIM), operating at {approx}1200 C, with selected HLW (AZ-101) and LAW (AN-105) successfully developed glasses with significant increase of waste loading compared to that is likely to be achieved based on expected reference WTP formulations. Three glasses formulated for AZ-101HLW and one glass for AN-105 LAW were selected for the initial CCIM demonstration melter tests. Melter tests were not performed within the finished scope of the WP-4.1.2 task. Glass formulations for CCIM were expanded to cover additional HLWs that have high potential to successfully demonstrate the unique advantages of the CCIM technologies based on projected composition of Hanford wastes. However, only the preliminary scoping tests were completed with selected wastes within the finished scope. Advanced glass formulations for the reference WTP melter, operating at {approx}1200 C, were initiated with selected specific wastes to determine the estimated maximum waste loading. The incomplete results from these initial formulation efforts are summarized. For systematic glass studies, a test matrix of 32 high-aluminum glasses was completed based on a new method developed in this study.

  4. 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.

    2013-06-01T23:59:59.000Z

    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.

  5. Precipitate hydrolysis process for the removal of organic compounds from nuclear waste slurries

    DOE Patents [OSTI]

    Doherty, J.P.; Marek, J.C.

    1987-02-25T23:59:59.000Z

    A process for removing organic compounds from a nuclear waste slurry comprising reacting a mixture of radioactive waste precipitate slurry and an acid in the presence of a catalytically effective amount of a copper(II) catalyst whereby the organic compounds in the precipitate slurry are hydrolyzed to form volatile organic compounds which are separated from the reacting mixture. The resulting waste slurry, containing less than 10 percent of the original organic compounds, is subsequently blended with high level radioactive sludge land transferred to a vitrification facility for processing into borosilicate glass for long-term storage. 2 figs., 3 tabs.

  6. Technical work plan for the privatization waste characterization data quality objective process

    SciTech Connect (OSTI)

    Kirkbride, R.A.

    1996-04-24T23:59:59.000Z

    This work plan addresses the activities necessary to complete the data quality objectives process for the purpose of providing sufficient characterization information to successfully stage, pretreat, and immobilize low-activity waste per the requirements and specifications identified in the Tank Waste Remediation System Privatization Request for Proposal. The scope of this task is to complete the data quality objectives process, the results of which will provide a technical basis for sampling and characterization needs related to privatization of pretreatment and low-activity waste immobilization.

  7. Precipitate hydrolysis process for the removal of organic compounds from nuclear waste slurries

    DOE Patents [OSTI]

    Doherty, Joseph P. (Elkton, MD); Marek, James C. (Augusta, GA)

    1989-01-01T23:59:59.000Z

    A process for removing organic compounds from a nuclear waste slurry comprising reacting a mixture of radioactive waste precipitate slurry and an acid in the presence of a catalytically effective amount of a copper (II) catalyst whereby the organic compounds in the precipitate slurry are hydrolyzed to form volatile organic compounds which are separated from the reacting mixture. The resulting waste slurry, containing less than 10 percent of the orginal organic compounds, is subsequently blended with high level radioactive sludge and transferred to a virtrification facility for processing into borosilicate glass for long-term storage.

  8. Waste Receiving and Processing Facility Module 2A: Advanced Conceptual Design Report. Volume 2

    SciTech Connect (OSTI)

    Not Available

    1994-03-01T23:59:59.000Z

    This volume presents the Total Estimated Cost (TEC) for the WRAP (Waste Receiving and Processing) 2A facility. The TEC is $81.9 million, including an overall project contingency of 25% and escalation of 13%, based on a 1997 construction midpoint. (The mission of WRAP 2A is to receive, process, package, certify, and ship for permanent burial at the Hanford site disposal facilities the Category 1 and 3 contact handled low-level radioactive mixed wastes that are currently in retrievable storage, and are forecast to be generated over the next 30 years by Hanford, and waste to be shipped to Hanford site from about 20 DOE sites.)

  9. Experimental study of the atomization process for viscous liquids by meniscus perturbation and micro air jet

    E-Print Network [OSTI]

    Lee, Heejin, 1976-

    2004-01-01T23:59:59.000Z

    An experimental study was performed to understand the atomization process in the specialized printing method which consists of the high frequency oscillating motion of a needle and a micro air jet. Highly viscous liquids, ...

  10. Liquid-phase Processing of Fast Pyrolysis Bio-oil using Pt/HZSM-5 Catalyst 

    E-Print Network [OSTI]

    Santos, Bjorn Sanchez

    2013-05-01T23:59:59.000Z

    Recent developments in converting biomass to bio-chemicals and liquid fuels provide a promising sight to an emerging biofuels industry. Biomass can be converted to energy via thermochemical and biochemical pathways. Thermal degradation processes...

  11. Environmental impacts of unmanaged solid waste at a former base metal mining and ore processing site (Kirki, Greece)

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Environmental impacts of unmanaged solid waste at a former base metal mining and ore processing-00553648,version1-30Jan2014 Author manuscript, published in "Waste Management & Research 28, 11 (2010) p, among the mining waste abandoned at a mine and processing plant, the most critical potential pollution

  12. Municipal Solid Waste (MSW) to Liquid Fuels Synthesis, Volume 2: A Techno-economic Evaluation of the Production of Mixed Alcohols

    SciTech Connect (OSTI)

    Jones, Susanne B.; Zhu, Yunhua; Valkenburt, Corinne

    2009-05-01T23:59:59.000Z

    Biomass is a renewable energy resource that can be converted into liquid fuel suitable for transportation applications and thus help meet the Energy Independence and Security Act renewable energy goals (U.S. Congress 2007). However, biomass is not always available in sufficient quantity at a price compatible with fuels production. Municipal solid waste (MSW) on the other hand is readily available in large quantities in some communities and is considered a partially renewable feedstock. Furthermore, MSW may be available for little or no cost. This report provides a techno-economic analysis of the production of mixed alcohols from MSW and compares it to the costs for a wood based plant. In this analysis, MSW is processed into refuse derived fuel (RDF) and then gasified in a plant co-located with a landfill. The resulting syngas is then catalytically converted to mixed alcohols. At a scale of 2000 metric tons per day of RDF, and using current technology, the minimum ethanol selling price at a 10% rate of return is approximately $1.85/gallon ethanol (early 2008 $). However, favorable economics are dependent upon the toxicity characteristics of the waste streams and that a market exists for the by-product scrap metal recovered from the RDF process.

  13. "Big Picture" Process Modeling Tools |GE Global Research

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

    downstream advantages as costs are reduced thanks to lower power requirements for distillation and even reduced liquid waste generation. In the second case, process modeling...

  14. Recovery of Valuable Chlorosilane Intermediates by a Novel Waste Conversion Process

    SciTech Connect (OSTI)

    J. Ashley Brinson

    2002-06-20T23:59:59.000Z

    From 1994 to 2001, Dow Corning studied a waste recycling process to recover direct process residues (DPR) resulting from the production of silicone precursors. Over the course of eight years, Dow Corning constructed and operated a pilot plant, a small scale commercial plant, and a full scale plant. The process reacts DPR with hydrogen and chlorosilane monomers at high temperature and high pressure. The process converted 85% of the DPR to valuable chlorosilane monomers such as dimethyldichlorosilane and methyldichlorosilane. When feeding methyltrichlorosilane, the process converted 30% of the MeSiCl3 to other monomers. Alternate co-feed monomers were tested. By converting waste DPR to valuable intermediates, the technology significantly reduces waste from the basic silicones manufacturing process.

  15. Pervaporation process and use in treating waste stream from glycol dehydrator

    DOE Patents [OSTI]

    Kaschemekat, Jurgen (Campbell, CA); Baker, Richard W. (Palo Alto, CA)

    1994-01-01T23:59:59.000Z

    Pervaporation processes and apparatus with few moving parts. Ideally, only one pump is used to provide essentially all of the motive power and driving force needed. The process is particularly useful for handling small streams with flow rates less than about 700 gpd. Specifically, the process can be used to treat waste streams from glycol dehydrator regeneration units.

  16. Waste Receiving and Processing (WRAP) Facility Final Safety Analysis Report (FSAR)

    SciTech Connect (OSTI)

    TOMASZEWSKI, T.A.

    2000-04-25T23:59:59.000Z

    The Waste Receiving and Processing Facility (WRAP), 2336W Building, on the Hanford Site is designed to receive, confirm, repackage, certify, treat, store, and ship contact-handled transuranic and low-level radioactive waste from past and present U.S. Department of Energy activities. The WRAP facility is comprised of three buildings: 2336W, the main processing facility (also referred to generically as WRAP); 2740W, an administrative support building; and 2620W, a maintenance support building. The support buildings are subject to the normal hazards associated with industrial buildings (no radiological materials are handled) and are not part of this analysis except as they are impacted by operations in the processing building, 2336W. WRAP is designed to provide safer, more efficient methods of handling the waste than currently exist on the Hanford Site and contributes to the achievement of as low as reasonably achievable goals for Hanford Site waste management.

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

    SciTech Connect (OSTI)

    Crawford, T.W.

    1995-09-22T23:59:59.000Z

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

  18. Hydrotreating process kinetics for bitumen and bitumen-derived liquids

    SciTech Connect (OSTI)

    Kwak, S.; Longstaff, D.C.; Deo, M.D.; Hanson, F.V.

    1993-01-01T23:59:59.000Z

    Hydrodenitrogenation, hydrodesulfurization and resid conversion data for the Whiterocks bitumen and bitumen-derived liquid were analyzed using a modified power rate law model. The model incorporated the space velocity and pressure since the plug flow equation may not be applicable to laboratory-scale reactors in which complete wetting of the catalyst may not be attained. The data were obtained with the reactor operating as a fixed bed reactor in the upflow mode. The space velocity (WHSV[sup [alpha

  19. US Department of Energy mixed waste characterization, treatment, and disposal focus area technical baseline development process

    SciTech Connect (OSTI)

    Roach, J.A.; Gombert, D. [Lockheed Martin Idaho Technologies, Idaho Falls, ID (United States)

    1996-12-31T23:59:59.000Z

    The US Department of Energy (DOE) created the Mixed Waste Characterization, Treatment, and Disposal Focus Area (MWFA) to develop and facilitate implementation of technologies required to meet its commitments for treatment of mixed wastes under the Federal Facility Compliance Act (FFCA), and in accordance with the Land Disposal Restrictions (LDR) of the Resource Conservation and Recovery Act (RCRA). Mixed wastes include both mixed low-level waste (MLLW) and mixed transuranic (MTRU) waste. The goal of the MWFA is to develop mixed waste treatment systems to the point of implementation by the Environmental Management (EM) customer. To accomplish this goal, the MWFA is utilizing a three step process. First, the treatment system technology deficiencies were identified and categorized. Second, these identified needs were prioritized. This resulted in a list of technical deficiencies that will be used to develop a technical baseline. The third step, the Technical Baseline Development Process, is currently ongoing. When finalized, the technical baseline will integrate the requirements associated with the identified needs into the planned and ongoing environmental research and technology development activities supported by the MWFA. Completion of this three-step process will result in a comprehensive technology development program that addresses customer identified and prioritized needs. The MWFA technical baseline will be a cost-effective, technically-defensible tool for addressing and resolving DOE`s mixed waste problems.

  20. Alcohol-free alkoxide process for containing nuclear waste

    DOE Patents [OSTI]

    Pope, James M. (Monroeville, PA); Lahoda, Edward J. (Edgewood, PA)

    1984-01-01T23:59:59.000Z

    Disclosed is a method of containing nuclear waste. A composition is first prepared of about 25 to about 80%, calculated as SiO.sub.2, of a partially hydrolyzed silicon compound, up to about 30%, calculated as metal oxide, of a partially hydrolyzed aluminum or calcium compound, about 5 to about 20%, calculated as metal oxide, of a partially hydrolyzed boron or calcium compound, about 3 to about 25%, calculated as metal oxide, of a partially hydrolyzed sodium, potassium or lithium compound, an alcohol in a weight ratio to hydrolyzed alkoxide of about 1.5 to about 3% and sufficient water to remove at least 99% of the alcohol as an azeotrope. The azeotrope is boiled off and up to about 40%, based on solids in the product, of the nuclear waste, is mixed into the composition. The mixture is evaporated to about 25 to about 45% solids and is melted and cooled.

  1. Formic Acid Free Flowsheet Development To Eliminate Catalytic Hydrogen Generation In The Defense Waste Processing

    SciTech Connect (OSTI)

    Lambert, Dan P.; Stone, Michael E.; Newell, J. David; Fellinger, Terri L.; Bricker, Jonathan M.

    2012-09-14T23:59:59.000Z

    The Defense Waste Processing Facility (DWPF) processes legacy nuclear waste generated at the Savannah River Site (SRS) during production of plutonium and tritium demanded by the Cold War. The nuclear waste is first treated via a complex sequence of controlled chemical reactions and then vitrified into a borosilicate glass form and poured into stainless steel canisters. Converting the nuclear waste into borosilicate glass canisters is a safe, effective way to reduce the volume of the waste and stabilize the radionuclides. Testing was initiated to determine whether the elimination of formic acid from the DWPF's chemical processing flowsheet would eliminate catalytic hydrogen generation. Historically, hydrogen is generated in chemical processing of alkaline High Level Waste sludge in DWPF. In current processing, sludge is combined with nitric and formic acid to neutralize the waste, reduce mercury and manganese, destroy nitrite, and modify (thin) the slurry rheology. The noble metal catalyzed formic acid decomposition produces hydrogen and carbon dioxide. Elimination of formic acid by replacement with glycolic acid has the potential to eliminate the production of catalytic hydrogen. Flowsheet testing was performed to develop the nitric-glycolic acid flowsheet as an alternative to the nitric-formic flowsheet currently being processed at the DWPF. This new flowsheet has shown that mercury can be reduced and removed by steam stripping in DWPF with no catalytic hydrogen generation. All processing objectives were also met, including greatly reducing the Slurry Mix Evaporator (SME) product yield stress as compared to the baseline nitric/formic flowsheet. Ten DWPF tests were performed with nonradioactive simulants designed to cover a broad compositional range. No hydrogen was generated in testing without formic acid.

  2. Re-refining of Waste Oil Solvent Is Used in Treatment/Distillation Process

    E-Print Network [OSTI]

    unknown authors

    INDUSTRIAL APPLICATION. A combination solvent treatment/distillation process has been designed for re-refining industrial waste oil (such as equipment lubricants, metal-working oil, and process oil) and used automotive lubricants (engine oil, hydraulic oil, and gear oil) for reuse. WASTE ENERGY RECOVERY. Recycling of waste oil in the United States has the potential to save the energy equivalent of 7-12 million bbl of crude oil annually.1 WASTE OIL RECOVERY. Prior to 1960, a significant portion of the demand for automotive lubricating oil was met by re-relined used oil. At the time, 150 re-refineries produced 300 million gal of motor oil annually. Since 1960, however, the production of re-refined oil has steadily declined. In 1981, for example, out of about 1.2 billion gal of automobile lubricating oil and 1.6 billion gal of industrial lubricating oils purchased, 25 U.S. rerefineries

  3. Hydrotreating process kinetics for bitumen and bitumen-derived liquids

    SciTech Connect (OSTI)

    Kwak, S.; Longstaff, D.C.; Deo, M.D.; Hanson, F.V.

    1993-03-01T23:59:59.000Z

    Hydrodenitrogenation, hydrodesulfurization and resid conversion data for the Whiterocks bitumen and bitumen-derived liquid were analyzed using a modified power rate law model. The model incorporated the space velocity and pressure since the plug flow equation may not be applicable to laboratory-scale reactors in which complete wetting of the catalyst may not be attained. The data were obtained with the reactor operating as a fixed bed reactor in the upflow mode. The space velocity (WHSV{sup {alpha}}) term was included to account for deviations from plug flow behavior. The exponents (a,p) and the kinetic parameters were obtained by combined non-linear regression and ODE solver techniques for the analysis of laboratory data. A simple nth order power rate law expression for hydrodenitrogenation and hydrodesulfurization was examined. The higher than first order kinetics for hydrodenitrogenation and hydrodesulfurization of the bitumen and bitumen-derived liquids were explained by invoking two parallel first-order reactions; one slow and the other fast. Parallel and consecutive reaction schemes were used to examine the extent of conversion of the resid fraction to middle distillate, gas oil and gasoline and the apparent kinetic parameters were determined. It was determined that the upflow operating mode was preferred to the trickle-bed mode in the laboratory reactor to insure plug flow behavior.

  4. Tank waste remediation system privatization infrastructure program requirements and document management process guide

    SciTech Connect (OSTI)

    ROOT, R.W.

    1999-05-18T23:59:59.000Z

    This guide provides the Tank Waste Remediation System Privatization Infrastructure Program management with processes and requirements to appropriately control information and documents in accordance with the Tank Waste Remediation System Configuration Management Plan (Vann 1998b). This includes documents and information created by the program, as well as non-program generated materials submitted to the project. It provides appropriate approval/control, distribution and filing systems.

  5. Review of Catalytic Hydrogen Generation in the Defense Waste Processing Facility (DWPF) Chemical Processing Cell

    SciTech Connect (OSTI)

    Koopman, D. C.

    2004-12-31T23:59:59.000Z

    This report was prepared to fulfill the Phase I deliverable for HLW/DWPF/TTR-98-0018, Rev. 2, ''Hydrogen Generation in the DWPF Chemical Processing Cell'', 6/4/2001. The primary objective for the preliminary phase of the hydrogen generation study was to complete a review of past data on hydrogen generation and to prepare a summary of the findings. The understanding was that the focus should be on catalytic hydrogen generation, not on hydrogen generation by radiolysis. The secondary objective was to develop scope for follow-up experimental and analytical work. The majority of this report provides a summary of past hydrogen generation work with radioactive and simulated Savannah River Site (SRS) waste sludges. The report also includes some work done with Hanford waste sludges and simulants. The review extends to idealized systems containing no sludge, such as solutions of sodium formate and formic acid doped with a noble metal catalyst. This includes general information from the literature, as well as the focused study done by the University of Georgia for the SRS. The various studies had a number of points of universal agreement. For example, noble metals, such as Pd, Rh, and Ru, catalyze hydrogen generation from formic acid and formate ions, and more acid leads to more hydrogen generation. There were also some points of disagreement between different sources on a few topics such as the impact of mercury on the noble metal catalysts and the identity of the most active catalyst species. Finally, there were some issues of potential interest to SRS that apparently have not been systematically studied, e.g. the role of nitrite ion in catalyst activation and reactivity. The review includes studies covering the period from about 1924-2002, or from before the discovery of hydrogen generation during simulant sludge processing in 1988 through the Shielded Cells qualification testing for Sludge Batch 2. The review of prior studies is followed by a discussion of proposed experimental work, additional data analysis, and future modeling programs. These proposals have led to recent investigations into the mercury issue and the effect of co-precipitating noble metals which will be documented in two separate reports. SRS hydrogen generation work since 2002 will also be collected and summarized in a future report on the effect of noble metal-sludge matrix interactions on hydrogen generation. Other potential factors for experimental investigation include sludge composition variations related to both the washing process and to the insoluble species with particular attention given to the role of silver and to improving the understanding of the interaction of nitrite ion with the noble metals.

  6. Separation and Purification and Beta Liquid Scintillation Analysis of Sm-151 in Savannah River Site and Hanford Site DOE High Level Waste

    SciTech Connect (OSTI)

    Dewberry, R.A.

    2001-02-13T23:59:59.000Z

    This paper describes development work to obtain a product phase of Sm-151 pure of any other radioactive species so that it can be determined in US Department of Energy high level liquid waste and low level solid waste by liquid scintillation {beta}-spectroscopy. The technique provides separation from {mu}Ci/ml levels of Cs-137, Pu alpha and Pu-241 {beta}-decay activity, and Sr-90/Y-90 activity. The separation technique is also demonstrated to be useful for the determination of Pm-147.

  7. Engineering assessment of low-level liquid waste disposal caisson locations at the 618-11 Burial Grounds

    SciTech Connect (OSTI)

    Phillips, S.J.; Fischer, D.D.; Crawford, R.C. [Westinghouse Hanford Co., Richland, WA (United States); Rising, J.L. [Pacific Northwest Lab., Richland, WA (United States)

    1982-06-01T23:59:59.000Z

    Rockwell Hanford Operations is currently involved in an extensive effort to perform interim ground surface stabilization activities at retired low-level waste burial grounds located at the Hanford Site, Richland, Washington. The principal objective of these activities is to promote increased occupational and radiological safety at burial grounds. Interim stabilization activities include: (1) load testing (traversing burial ground surfaces with heavy equipment to promote incipient collapse of void spaces within the disposal structure and overburden), (2) barrier placement (placement of a {ge} 0.6 m soil barrier over existing overburden), and (3) revegetation (establishment of shallow rooted vegetation on the barrier to mitigate deep rooted plant growth and to reduce erosion). Low-level waste disposal caissons were used in 300 Area Burial Grounds as internment structures for containerized liquid wastes. These caissons, by virtue of their contents, design and methods of closure, require long-term performance evaluation. As an initial activity to evaluate long-term performance, the accurate location of these structures is required. This topical report summarizes engineering activities used to locate caissons in the subsurface environment at the Burial Ground. Activities were conducted to locate caissons during surface stabilization activities. The surface locations were marked, photographed, and recorded on an as built engineering drawing. The recorded location of these caissons will augment long-term observations of confinement structure and engineered surface barrier performance. In addition, accurate caisson location will minimize occupational risk during monitoring and observation activities periodically conducted at the burial ground.

  8. Salt Waste Processing Facility Fact Sheet | Department of Energy

    Energy Savers [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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directed offOCHCO Overview OCHCO OverviewRepository | DepartmentSEA-04:Department of¿QUÉFuture |Waste

  9. Savannah River Site - Salt Waste Processing Facility Independent Technical Review

    Office of Environmental Management (EM)

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM615_CostNSARDevelopmentalEfficiency | Department ofEnergySaraKSALT WASTE

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

    SciTech Connect (OSTI)

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

    2010-11-01T23:59:59.000Z

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

  11. Laboratory Demonstration of the Pretreatment Process with Caustic and Oxidative Leaching Using Actual Hanford Tank Waste

    SciTech Connect (OSTI)

    Fiskum, Sandra K.; Billing, Justin M.; Buck, Edgar C.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Jenson, Evan D.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Shimskey, Rick W.; Snow, Lanee A.

    2009-01-01T23:59:59.000Z

    This report describes the bench-scale pretreatment processing of actual tank waste materials through the entire baseline WTP pretreatment flowsheet in an effort to demonstrate the efficacy of the defined leaching processes on actual Hanford tank waste sludge and the potential impacts on downstream pretreatment processing. The test material was a combination of reduction oxidation (REDOX) tank waste composited materials containing aluminum primarily in the form of boehmite and dissolved S saltcake containing Cr(III)-rich entrained solids. The pretreatment processing steps tested included • caustic leaching for Al removal • solids crossflow filtration through the cell unit filter (CUF) • stepwise solids washing using decreasing concentrations of sodium hydroxide with filtration through the CUF • oxidative leaching using sodium permanganate for removing Cr • solids filtration with the CUF • follow-on solids washing and filtration through the CUF • ion exchange processing for Cs removal • evaporation processing of waste stream recycle for volume reduction • combination of the evaporated product with dissolved saltcake. The effectiveness of each process step was evaluated by following the mass balance of key components (such as Al, B, Cd, Cr, Pu, Ni, Mn, and Fe), demonstrating component (Al, Cr, Cs) removal, demonstrating filterability by evaluating filter flux rates under various processing conditions (transmembrane pressure, crossflow velocities, wt% undissolved solids, and PSD) and filter fouling, and identifying potential issues for WTP. The filterability was reported separately (Shimskey et al. 2008) and is not repeated herein.

  12. Combustion process for synthesis of carbon nanomaterials from liquid hydrocarbon

    DOE Patents [OSTI]

    Diener, Michael D.; Alford, J. Michael; Nabity, James; Hitch, Bradley D.

    2007-01-02T23:59:59.000Z

    The present invention provides a combustion apparatus for the production of carbon nanomaterials including fullerenes and fullerenic soot. Most generally the combustion apparatus comprises one or more inlets for introducing an oxygen-containing gas and a hydrocarbon fuel gas in the combustion system such that a flame can be established from the mixed gases, a droplet delivery apparatus for introducing droplets of a liquid hydrocarbon feedstock into the flame, and a collector apparatus for collecting condensable products containing carbon nanomaterials that are generated in the combustion system. The combustion system optionally has a reaction zone downstream of the flame. If this reaction zone is present the hydrocarbon feedstock can be introduced into the flame, the reaction zone or both.

  13. State-of-the-art of liquid waste disposal for geothermal energy systems: 1979. Report PNL-2404

    SciTech Connect (OSTI)

    Defferding, L.J.

    1980-06-01T23:59:59.000Z

    The state-of-the-art of geothermal liquid waste disposal is reviewed and surface and subsurface disposal methods are evaluated with respect to technical, economic, legal, and environmental factors. Three disposal techniques are currently in use at numerous geothermal sites around the world: direct discharge into surface waters; deep-well injection; and ponding for evaporation. The review shows that effluents are directly discharged into surface waters at Wairakei, New Zealand; Larderello, Italy; and Ahuachapan, El Salvador. Ponding for evaporation is employed at Cerro Prieto, Mexico. Deep-well injection is being practiced at Larderello; Ahuachapan; Otake and Hatchobaru, Japan; and at The Geysers in California. All sites except Ahuachapan (which is injecting only 30% of total plant flow) have reported difficulties with their systems. Disposal techniques used in related industries are also reviewed. The oil industry's efforts at disposal of large quantities of liquid effluents have been quite successful as long as the effluents have been treated prior to injection. This study has determined that seven liquid disposal methods - four surface and three subsurface - are viable options for use in the geothermal energy industry. However, additional research and development is needed to reduce the uncertainties and to minimize the adverse environmental impacts of disposal. (MHR)

  14. The thermodynamics of pyrochemical processes for liquid metal reactor fuel cycles

    SciTech Connect (OSTI)

    Johnson, I.

    1987-01-01T23:59:59.000Z

    The thermodynamic basis for pyrochemical processes for the recovery and purification of fuel for the liquid metal reactor fuel cycle is described. These processes involve the transport of the uranium and plutonium from one liquid alloy to another through a molten salt. The processes discussed use liquid alloys of cadmium, zinc, and magnesium and molten chloride salts. The oxidation-reduction steps are done either chemically by the use of an auxiliary redox couple or electrochemically by the use of an external electrical supply. The same basic thermodynamics apply to both the salt transport and the electrotransport processes. Large deviations from ideal solution behavior of the actinides and lanthanides in the liquid alloys have a major influence on the solubilities and the performance of both the salt transport and electrotransport processes. Separation of plutonium and uranium from each other and decontamination from the more noble fission product elements can be achieved using both transport processes. The thermodynamic analysis is used to make process design computations for different process conditions.

  15. Method of drying passivated micromachines by dewetting from a liquid-based process

    DOE Patents [OSTI]

    Houston, Michael R. (Berkeley, CA); Howe, Roger T. (Lafayette, CA); Maboudian, Roya (Orinda, CA); Srinivasan, Uthara (Berkeley, CA)

    2000-01-01T23:59:59.000Z

    A method of fabricating a micromachine includes the step of constructing a low surface energy film on the micromachine. The micromachine is then rinsed with a rinse liquid that has a high surface energy, relative to the low surface energy film, to produce a contact angle of greater than 90.degree. between the low surface energy film and the rinse liquid. This relatively large contact angle causes any rinse liquid on the micromachine to be displaced from the micromachine when the micromachine is removed from the rinse liquid. In other words, the micromachine is dried by dewetting from a liquid-based process. Thus, a separate evaporative drying step is not required, as the micromachine is removed from the liquid-based process in a dry state. The relatively large contact angle also operates to prevent attractive capillary forces between micromachine components, thereby preventing contact and adhesion between adjacent microstructure surfaces. The low surface energy film may be constructed with a fluorinated self-assembled monolayer film. The processing of the invention avoids the use of environmentally harmful, health-hazardous chemicals.

  16. National Institutes of Health: Mixed waste minimization and treatment

    SciTech Connect (OSTI)

    NONE

    1995-08-01T23:59:59.000Z

    The Appalachian States Low-Level Radioactive Waste Commission requested the US Department of Energy`s National Low-Level Waste Management Program (NLLWMP) to assist the biomedical community in becoming more knowledgeable about its mixed waste streams, to help minimize the mixed waste stream generated by the biomedical community, and to identify applicable treatment technologies for these mixed waste streams. As the first step in the waste minimization process, liquid low-level radioactive mixed waste (LLMW) streams generated at the National Institutes of Health (NIH) were characterized and combined into similar process categories. This report identifies possible waste minimization and treatment approaches for the LLMW generated by the biomedical community identified in DOE/LLW-208. In development of the report, on site meetings were conducted with NIH personnel responsible for generating each category of waste identified as lacking disposal options. Based on the meetings and general waste minimization guidelines, potential waste minimization options were identified.

  17. Solid Waste Integrated Forecast Technical (SWIFT) Report FY2001 to FY2046 Volume 1

    SciTech Connect (OSTI)

    BARCOT, R.A.

    2000-08-31T23:59:59.000Z

    This report provides up-to-date life cycle information about the radioactive solid waste expected to be managed by Hanford's Waste Management (WM) Project from onsite and offsite generators. It includes: an overview of Hanford-wide solid waste to be managed by the WM Project; program-level and waste class-specific estimates; background information on waste sources; and comparisons to previous forecasts and other national data sources. This report does not include: waste to be managed by the Environmental Restoration (EM-40) contractor (i.e., waste that will be disposed of at the Environmental Restoration Disposal Facility (ERDF)); waste that has been received by the WM Project to date (i.e., inventory waste); mixed low-level waste that will be processed and disposed by the River Protection Program; and liquid waste (current or future generation). Although this report currently does not include liquid wastes, they may be added as information becomes available.

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

    SciTech Connect (OSTI)

    Bonnema, Bruce Edward

    2001-09-01T23:59:59.000Z

    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.

  19. Conditioning matrices from high level waste resulting from pyrochemical processing in fluorine salt

    SciTech Connect (OSTI)

    Grandjean, Agnes; Advocat, Thierry; Bousquet, Nicolas [SCDV - Service de Confinement des Dechets et Vitrification - Laboratoire d'Etudes de Base sur les Verres, CEA Valrho, Centre de Marcoule, 30207 Bagnols sur Ceze (France); Jegou, Christophe [SECM - Service d'Etude du Confinement et Materiaux - Laboratoire des Materiaux et Procedes Actifs - CEA Valrho, Centre de Marcoule, 30207 Bagnols sur Ceze (France)

    2007-07-01T23:59:59.000Z

    Separating the actinides from the fission products through reductive extraction by aluminium in a LiF/AlF{sub 3} medium is a process investigated for pyrometallurgical reprocessing of spent fuel. The process involves separation by reductive salt-metal extraction. After dissolving the fuel or the transmutation target in a salt bath, the noble metal fission products are first extracted by contacting them with a slightly reducing metal. After extracting the metal fission products, then the actinides are selectively separated from the remaining fission products. In this hypothesis, all the unrecoverable fission products would be conditioned as fluorides. Therefore, this process will generate first a metallic waste containing the 'reducible' fission products (Pd, Mo, Ru, Rh, Tc, etc.) and a fluorine waste containing alkali-metal, alkaline-earth and rare earth fission products. Immobilization of these wastes in classical borosilicate glasses is not feasible due to the very low solubility of noble metals, and of fluoride in these hosts. Alternative candidates have therefore been developed including silicate glass/ceramic system for fluoride fission products and metallic ones for noble metal fission products. These waste-forms were evaluated for their confinement properties like homogeneity, waste loading, volatility during the elaboration process, chemical durability, etc. using appropriate techniques. (authors)

  20. Savannah River Site - Salt Waste Processing Facility: Briefing on the Salt Waste Processing Facility Independent Technical Review

    Office of Environmental Management (EM)

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM615_CostNSARDevelopmentalEfficiency | Department ofEnergySaraKSALT WASTE

  1. CONDITIONING TECHNOLOGY FOR RADIOACTIVE WASTE RESULTED FROM THE TREATMENT OF LIQUID WASTE FROM THE ROMANIAN NUCLEAR POWER PLANT

    SciTech Connect (OSTI)

    ARSENE, CARMEN; ANDREI, VERONICA; NEGOIU, DUMITRU

    2003-02-27T23:59:59.000Z

    For the conditioning of spent resins contaminated with radionuclides, such as: 137Cs, 134Cs, 60Co, 58Co, 57Co, 54Mn, etc., techniques of direct immobilization in cement, bitumen and organic polymers have been tested. The selected process was the bituminization using industrial bitumen, I 60-70, made in Romania, which had very good characteristics. The paper presents stages of the research project, technical conditions for the process and advantages of the bituminization of spent resins.

  2. CHARACTERIZATION OF INDIVIDUAL CHEMICAL REACTIONS CONSUMING ACID DURING NUCLEAR WASTE PROCESSING AT THE SAVANNAH RIVER SITE - 136B

    SciTech Connect (OSTI)

    Koopman, D.; Pickenheim, B.; Lambert, D.; Newell, J.; Stone, M.

    2009-09-02T23:59:59.000Z

    Conversion of legacy radioactive high-level waste at the Savannah River Site into a stable glass waste form involves a chemical pretreatment process to prepare the waste for vitrification. Waste slurry is treated with nitric and formic acids to achieve certain goals. The total quantity of acid added to a batch of waste slurry is constrained by the catalytic activity of trace noble metal fission products in the waste that can convert formic acid into hydrogen gas at many hundreds of times the radiolytic hydrogen generation rate. A large block of experimental process simulations were performed to characterize the chemical reactions that consume acid prior to hydrogen generation. The analysis led to a new equation for predicting the quantity of acid required to process a given volume of waste slurry.

  3. Application of advanced liquid metal reactors to the destruction of radioactive wastes

    SciTech Connect (OSTI)

    Karnesky, R.A.; Dobbin, K.D.; Jordheim, D.P.; Rawlins, J.A.; Wootan, D.W.

    1991-08-01T23:59:59.000Z

    The ability of a small fast reactor to destroy hazardous long lived minor actinide and fission product wastes is evaluated. It is determined that by using a novel technique wherein high energy neutrons leaking from the active core of the reactor are moderated by yttrium hydride located in the target assemblies, substantial amounts of long lived fission products can be destroyed and useful quantities of the beneficial isotope {sup 238}Pu can be produced by transmutation of the {sup 237}Np and {sub 241}Am minor actinide waste components. In addition it is shown that minor actinides recovered from spent Light Water Reactor fuel can be used to fuel such a reactor, increasing the amount of hazardous minor actinide and fission product wastes that can be destroyed. 9 refs., 4 figs., 2 tabs.

  4. Development of Vitrification Process and Glass Formulation for Nuclear Waste Conditioning

    SciTech Connect (OSTI)

    Petitjean, V.; Fillet, C.; Boen, R.; Veyer, C.; Flament, T.

    2002-02-26T23:59:59.000Z

    The vitrification of high-level waste is the internationally recognized standard to minimize the impact to the environment resulting from waste disposal as well as to minimize the volume of conditioned waste to be disposed of. COGEMA has been vitrifying high-level waste industrially for over 20 years and is currently operating three commercial vitrification facilities based on a hot metal crucible technology, with outstanding records of safety, reliability and product quality. To further increase the performance of vitrification facilities, CEA and COGEMA have been developing the cold crucible melter technology since the beginning of the 1980s. This type of melter is characterized by a virtually unlimited equipment service life and a great flexibility in dealing with various types of waste and allowing development of high temperature matrices. In complement of and in parallel with the vitrification process, a glass formulation methodology has been developed by the CEA in order to tailor matrices for the wastes to be conditioned while providing the best adaptation to the processing technology. The development of a glass formulation is a trade-off between material properties and qualities, technical feasibility, and disposal safety criteria. It involves non-radioactive and radioactive laboratories in order to achieve a comprehensive matrix qualification. Several glasses and glass ceramics have thus been studied by the CEA to be compliant with industrial needs and waste characteristics: glasses or other matrices for a large spectrum of fission products, or for high contents of specifics elements such as sodium, phosphate, iron, molybdenum, or actinides. New glasses or glass-ceramics designed to minimize the final wasteform volume for solutions produced during the reprocessing of high burnup fuels or to treat legacy wastes are now under development and take benefit from the latest CEA hot-laboratories and technology development. The paper presents the CEA state-of-the-art in developing matrices or glasses and provides several examples.

  5. Dilution and dispersion of liquid wastes in the Gulf of Mexico

    E-Print Network [OSTI]

    Beckett, Daniel Edward

    1974-01-01T23:59:59.000Z

    Sanctuaries Act of 1972, however, the Environmental Protection Agency (EPA) has been designated to oversee the barging of these wastes to sea. Thus, before a discharger can legally discharge in the ocean, he must acquire a permit from the EPA to do so. Data... legislation to provide any form of regulation of waste disposal practices. The Rivers and Harbors Act of 1899 prohi- bited the dumping of unauthorized materials into the navigable waters of the United States. This Act was originally designed to prevent...

  6. Corrosion Control Measures For Liquid Radioactive Waste Storage Tanks At The Savannah River Site

    SciTech Connect (OSTI)

    Wiersma, B. J.; Subramanian, K. H.

    2012-11-27T23:59:59.000Z

    The Savannah River Site has stored radioactive wastes in large, underground, carbon steel tanks for approximately 60 years. An assessment of potential degradation mechanisms determined that the tanks may be vulnerable to nitrate- induced pitting corrosion and stress corrosion cracking. Controls on the solution chemistry and temperature of the wastes are in place to mitigate these mechanisms. These controls are based upon a series of experiments performed using simulated solutions on materials used for construction of the tanks. The technical bases and evolution of these controls is presented in this paper.

  7. Review of the Savannah River Site, Salt Waste Processing Facility...

    Energy Savers [EERE]

    Grade Dedication and Material Upgrade Package CMTR Certified Material Test Report CPA Central Process Area COC Certificate of Conformance CR Condition Report CRAD Criteria and...

  8. Rotor assembly and method for automatically processing liquids

    DOE Patents [OSTI]

    Burtis, C.A.; Johnson, W.F.; Walker, W.A.

    1992-12-22T23:59:59.000Z

    A rotor assembly is described for performing a relatively large number of processing steps upon a sample, such as a whole blood sample, and a diluent, such as water. It includes a rotor body for rotation about an axis and includes a network of chambers within which various processing steps are performed upon the sample and diluent and passageways through which the sample and diluent are transferred. A transfer mechanism is movable through the rotor body by the influence of a magnetic field generated adjacent the transfer mechanism and movable along the rotor body, and the assembly utilizes centrifugal force, a transfer of momentum and capillary action to perform any of a number of processing steps such as separation, aliquoting, transference, washing, reagent addition and mixing of the sample and diluent within the rotor body. The rotor body is particularly suitable for automatic immunoassay analyses. 34 figs.

  9. Rotor assembly and method for automatically processing liquids

    DOE Patents [OSTI]

    Burtis, Carl A. (Oak Ridge, TN); Johnson, Wayne F. (Loudon, TN); Walker, William A. (Knoxville, TN)

    1992-01-01T23:59:59.000Z

    A rotor assembly for performing a relatively large number of processing steps upon a sample, such as a whole blood sample, and a diluent, such as water, includes a rotor body for rotation about an axis and including a network of chambers within which various processing steps are performed upon the sample and diluent and passageways through which the sample and diluent are transferred. A transfer mechanism is movable through the rotor body by the influence of a magnetic field generated adjacent the transfer mechanism and movable along the rotor body, and the assembly utilizes centrifugal force, a transfer of momentum and capillary action to perform any of a number of processing steps such as separation, aliquoting, transference, washing, reagent addition and mixing of the sample and diluent within the rotor body. The rotor body is particularly suitable for automatic immunoassay analyses.

  10. Catalysts and process for liquid hydrocarbon fuel production

    DOE Patents [OSTI]

    White, Mark G; Liu, Shetian

    2014-12-09T23:59:59.000Z

    The present invention provides a novel process and system in which a mixture of carbon monoxide and hydrogen synthesis gas, or syngas, is converted into hydrocarbon mixtures composed of high quality gasoline components, aromatic compounds, and lower molecular weight gaseous olefins in one reactor or step. The invention utilizes a novel molybdenum-zeolite catalyst in high pressure hydrogen for conversion, as well as a novel rhenium-zeolite catalyst in place of the molybdenum-zeolite catalyst, and provides for use of the novel catalysts in the process and system of the invention.

  11. Ceramicrete stabilization of radioactive-salt-containing liquid waste and sludge water. Final CRADA report.

    SciTech Connect (OSTI)

    Ehst, D.; Nuclear Engineering Division

    2010-08-04T23:59:59.000Z

    It was found that the Ceramicrete Specimens incorporated the Streams 1 and 2 sludges with the adjusted loading about 41.6 and 31.6%, respectively, have a high solidity. The visible cracks in the matrix materials and around the anionite AV-17 granules included could not obtain. The granules mentioned above fixed by Ceramicrete matrix very strongly. Consequently, we can conclude that irradiation of Ceramecrete matrix, goes from the high radioactive elements, not result the structural degradation. Based on the chemical analysis of specimens No.462 and No.461 used it was shown that these matrix included the formation elements (P, K, Mg, O), but in the different samples their correlations are different. These ratios of the content of elements included are about {+-} 10%. This information shows a great homogeneity of matrix prepared. In the list of the elements founded, expect the matrix formation elements, we detected also Ca and Si (from the wollastonite - the necessary for Ceramicrete compound); Na, Al, S, O, Cl, Fe, Ni also have been detected in the Specimen No.642 from the waste forms: NaCl, Al(OH){sub 3}, Na{sub 2}SO{sub 4}. Fe(OH){sub 3}, nickel ferrocyanide and Ni(NO{sub 3})2. The unintelligible results also were found from analysis of an AV-17 granules, in which we obtain the great amount of K. The X-ray radiographs of the Ceramicrete specimens with loading 41.4 % of Stream 1 and 31.6% of Stream 2, respectively showed that the realization of the advance technology, created at GEOHKI, leads to formation of excellent ceramic matrix with high amount of radioactive streams up to 40% and more. Really, during the interaction with start compounds MgO and KH{sub 2}PO{sub 4} with the present of H{sub 3}BO{sub 3} and Wollastonite this process run with high speed under the controlled regimes. That fact that the Ceramicrete matrix with 30-40% of Streams 1 and 2 have a crystalline form, not amorphous matter, allows to permit that these matrix should be very stable, reliable for incorporation of a radionuclides.

  12. Non-Invasive Diagnostics for Measuring Physical Properties and Processes in High Level Wastes

    SciTech Connect (OSTI)

    Robert Powell; David Pfund

    2005-07-17T23:59:59.000Z

    This research demonstrated the usefulness of tomographic techniques for determining the physical properties of slurry suspensions. Of particular interest was the measurement of the viscosity of suspensions in complex liquids and modeling these. We undertook a long rage program that used two techniques, magnetic resonance imaging and ultrasonic pulsed Doppler velocimetry. Our laboratory originally developed both of these for the measurement of viscosity of complex liquids and suspensions. We have shown that the relationship between shear viscosity and shear rate can be determined over a wide range of shear rates from a single measurement. We have also demonstrated these techniques for many non-Newtonian fluids which demonstrate highly shear thinning behavior. This technique was extended to determine the yield stress with systems of interacting particles. To model complex slurries that may be found in wastes applications, we have also used complex slurries that are found in industrial applications

  13. Bioelectrochemical Integration of Waste Heat Recovery, Waste...

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

    Bioelectrochemical Integration of Waste Heat Recovery, Waste-to-Energy Conversion, and Waste-to-Chemical Conversion with Industrial Gas and Chemical Manufacturing Processes...

  14. Advanced Thermoelectric Materials for Efficient Waste Heat Recovery in Process Industries

    SciTech Connect (OSTI)

    Adam Polcyn; Moe Khaleel

    2009-01-06T23:59:59.000Z

    The overall objective of the project was to integrate advanced thermoelectric materials into a power generation device that could convert waste heat from an industrial process to electricity with an efficiency approaching 20%. Advanced thermoelectric materials were developed with figure-of-merit ZT of 1.5 at 275 degrees C. These materials were not successfully integrated into a power generation device. However, waste heat recovery was demonstrated from an industrial process (the combustion exhaust gas stream of an oxyfuel-fired flat glass melting furnace) using a commercially available (5% efficiency) thermoelectric generator coupled to a heat pipe. It was concluded that significant improvements both in thermoelectric material figure-of-merit and in cost-effective methods for capturing heat would be required to make thermoelectric waste heat recovery viable for widespread industrial application.

  15. DOE Awards Contract for Oak Ridge Transuranic Waste Processing...

    Energy Savers [EERE]

    Laboratory; and process Nuclear Fuel Services soils. The contract contains Firm-Fixed-Price (FFP) and Cost-Plus-Award-Fee (CPAF) Contract Line Items (CLINs), as well as an...

  16. Radioanalytical Chemistry for Automated Nuclear Waste Process Monitoring

    SciTech Connect (OSTI)

    Jay W. Grate; Timothy A. DeVol

    2006-07-20T23:59:59.000Z

    The objectives of our research were to develop the first automated radiochemical process analyzer including sample pretreatment methodoology, and to initiate work on new detection approaches, especially using modified diode detectors.

  17. Integrated process analysis of treatment systems for mixed low level waste

    SciTech Connect (OSTI)

    Cooley, C.R. [Dept. of Energy, Washington, DC (United States); Schwinkendorf, W.E. [Lockheed Martin Idaho Technology Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.]|[Sandia National Labs., Albuquerque, NM (United States); Bechtold, T.E. [Lockheed Martin Idaho Technology Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

    1997-10-01T23:59:59.000Z

    Selection of technologies to be developed for treatment of DOE`s mixed low level waste (MLLW) requires knowledge and understanding of the expected costs, schedules, risks, performance, and reliability of the total engineered systems that use these technologies. Thus, an integrated process analysis program was undertaken to identify the characteristics and needs of several thermal and nonthermal systems. For purposes of comparison, all systems were conceptually designed for a single facility processing the same amount of waste at the same rate. Thirty treatment systems were evaluated ranging from standard incineration to innovative thermal systems and innovative nonthermal chemical treatment. Treating 236 million pounds of waste in 20 years through a central treatment was found to be the least costly option with total life cycle cost ranging from $2.1 billion for a metal melting system to $3.9 billion for a nonthermal acid digestion system. Little cost difference exists among nonthermal systems or among thermal systems. Significant cost savings could be achieved by working towards maximum on line treatment time per year; vitrifying the final waste residue; decreasing front end characterization segregation and sizing requirements; using contaminated soil as the vitrifying agent; and delisting the final vitrified waste form from Resource Conservation and Recovery Act (RCRA) Land Disposal Restriction (LDR) requirements.

  18. Microwave applicator for in-drum processing of radioactive waste slurry

    DOE Patents [OSTI]

    White, Terry L. (Oak Ridge, TN)

    1994-01-01T23:59:59.000Z

    A microwave applicator for processing of radioactive waste slurry uses a waveguide network which splits an input microwave of TE.sub.10 rectangular mode to TE.sub.01 circular mode. A cylindrical body has four openings, each receiving 1/4 of the power input. The waveguide network includes a plurality of splitters to effect the 1/4 divisions of power.

  19. Project management plan, Waste Receiving and Processing Facility, Module 1, Project W-026

    SciTech Connect (OSTI)

    Starkey, J.G.

    1993-05-01T23:59:59.000Z

    The Hanford Waste Receiving and Processing Facility Module 1 Project (WRAP 1) has been established to support the retrieval and final disposal of approximately 400K grams of plutonium and quantities of hazardous components currently stored in drums at the Hanford Site.

  20. IWater Processing and Waste Management SystemsIntegrated System Health Management 2007 Phase II

    E-Print Network [OSTI]

    and other sources. It may serve as a backup processor for one or more existing water recovery systems unit drying prototype to for the recovery and recycle of water from concentrated waste water recovery system surfaces. The process is suitable for dewatering brine from the vapor compression distillation processor

  1. FULL SCALE TESTING TECHNOLOGY MATURATION OF A THIN FILM EVAPORATOR FOR HIGH-LEVEL LIQUID WASTE MANAGEMENT AT HANFORD - 12125

    SciTech Connect (OSTI)

    TEDESCHI AR; CORBETT JE; WILSON RA; LARKIN J

    2012-01-26T23:59:59.000Z

    Simulant testing of a full-scale thin-film evaporator system was conducted in 2011 for technology development at the Hanford tank farms. Test results met objectives of water removal rate, effluent quality, and operational evaluation. Dilute tank waste simulant, representing a typical double-shell tank supernatant liquid layer, was concentrated from a 1.1 specific gravity to approximately 1.5 using a 4.6 m{sup 2} (50 ft{sup 2}) heated transfer area Rototherm{reg_sign} evaporator from Artisan Industries. The condensed evaporator vapor stream was collected and sampled validating efficient separation of the water. An overall decontamination factor of 1.2E+06 was achieved demonstrating excellent retention of key radioactive species within the concentrated liquid stream. The evaporator system was supported by a modular steam supply, chiller, and control computer systems which would be typically implemented at the tank farms. Operation of these support systems demonstrated successful integration while identifying areas for efficiency improvement. Overall testing effort increased the maturation of this technology to support final deployment design and continued project implementation.

  2. Isolation of Metals from Liquid Wastes: Reactive Scavenging in Turbulent Thermal Reactors

    SciTech Connect (OSTI)

    Jost O.L. Wendt; Alan R. Kerstein; Alexander Scheeline; Arne Pearlstein; William Linak

    2003-08-06T23:59:59.000Z

    The Overall project demonstrated that toxic metals (cesium Cs and strontium Sr) in aqueous and organic wastes can be isolated from the environment through reaction with kaolinite based sorbent substrates in high temperature reactor environments. In addition, a state-of-the art laser diagnostic tool to measure droplet characteristic in practical 'dirty' laboratory environments was developed, and was featured on the cover of a recent edition of the scientific journal ''applied Spectroscopy''. Furthermore, great strides have been made in developing a theoretical model that has the potential to allow prediction of the position and life history of every particle of waste in a high temperature, turbulent flow field, a very challenging problem involving as it does, the fundamentals of two phase turbulence and of particle drag physics.

  3. Catalytic two-stage coal hydrogenation process using extinction recycle of heavy liquid fraction

    DOE Patents [OSTI]

    MacArthur, J.B.; Comolli, A.G.; McLean, J.B.

    1989-10-17T23:59:59.000Z

    A process is described for catalytic two-stage hydrogenation and liquefaction of coal with selective extinction recycle of all heavy liquid fractions boiling above a distillation cut point of about 600--750 F to produce increased yields of low-boiling hydrocarbon liquid and gas products. In the process, the particulate coal feed is slurried with a process-derived liquid solvent normally boiling above about 650 F and fed into a first stage catalytic reaction zone operated at conditions which promote controlled rate liquefaction of the coal, while simultaneously hydrogenating the hydrocarbon recycle oils. The first stage reactor is maintained at 710--800 F temperature, 1,000--4,000 psig hydrogen partial pressure, and 10-90 lb/hr per ft[sup 3] catalyst space velocity. Partially hydrogenated material withdrawn from the first stage reaction zone is passed directly to the second stage catalytic reaction zone maintained at 760--860 F temperature for further hydrogenation and hydroconversion reactions. A 600--750 F[sup +] fraction containing 0--20 W % unreacted coal and ash solids is recycled to the coal slurrying step. If desired, the cut point lower boiling fraction can be further catalytically hydrotreated. By this process, the coal feed is successively catalytically hydrogenated and hydroconverted at selected conditions, to provide significantly increased yields of desirable low-boiling hydrocarbon liquid products and minimal production of hydrocarbon gases, and no net production of undesirable heavy oils and residuum materials. 2 figs.

  4. Catalytic two-stage coal hydrogenation process using extinction recycle of heavy liquid fraction

    DOE Patents [OSTI]

    MacArthur, James B. (Denville, NJ); Comolli, Alfred G. (Yardley, PA); McLean, Joseph B. (Somerville, NJ)

    1989-01-01T23:59:59.000Z

    A process for catalytic two-stage hydrogenation and liquefaction of coal with selective extinction recycle of all heavy liquid fractions boiling above a distillation cut point of about 600.degree.-750.degree. F. to produce increased yields of low-boiling hydrocarbon liquid and gas products. In the process, the particulate coal feed is slurried with a process-derived liquid solvent normally boiling above about 650.degree. F. and fed into a first stage catalytic reaction zone operated at conditions which promote controlled rate liquefaction of the coal, while simultaneously hydrogenating the hydrocarbon recycle oils. The first stage reactor is maintained at 710.degree.-800.degree. F. temperature, 1000-4000 psig hydrogen partial pressure, and 10-90 lb/hr per ft.sup.3 catalyst space velocity. Partially hydrogenated material withdrawn from the first stage reaction zone is passed directly to the second stage catalytic reaction zone maintained at 760.degree.-860.degree. F. temperature for further hydrogenation and hydroconversion reactions. A 600.degree.-750.degree. F..sup.+ fraction containing 0-20 W % unreacted coal and ash solids is recycled to the coal slurrying step. If desired, the cut point lower boiling fraction can be further catalytically hydrotreated. By this process, the coal feed is successively catalytically hydrogenated and hydroconverted at selected conditions, to provide significantly increased yields of desirable low-boiling hydrocarbon liquid products and minimal production of hydrocarbon gases, and no net production of undesirable heavy oils and residuum materials.

  5. Process for removal of mineral particulates from coal-derived liquids

    DOE Patents [OSTI]

    McDowell, William J. (Knoxville, TN)

    1980-01-01T23:59:59.000Z

    Suspended mineral solids are separated from a coal-derived liquid containing the solids by a process comprising the steps of: (a) contacting said coal-derived liquid containing solids with a molten additive having a melting point of 100.degree.-500.degree. C. in an amount of up to 50 wt. % with respect to said coal-derived liquid containing solids, said solids present in an amount effective to increase the particle size of said mineral solids and comprising material or mixtures of material selected from the group of alkali metal hydroxides and inorganic salts having antimony, tin, lithium, sodium, potassium, magnesium, calcium, beryllium, aluminum, zinc, molybdenum, cobalt, nickel, ruthenium, rhodium or iron cations and chloride, iodide, bromide, sulfate, phosphate, borate, carbonate, sulfite, or silicate anions; and (b) maintaining said coal-derived liquid in contact with said molten additive for sufficient time to permit said mineral matter to agglomerate, thereby increasing the mean particle size of said mineral solids; and (c) recovering a coal-derived liquid product having reduced mineral solids content. The process can be carried out with less than 5 wt. % additive and in the absence of hydrogen pressure.

  6. Process model and capacity upgrades of the CTI-4000 liquid helium coldbox

    SciTech Connect (OSTI)

    Hansen, Benjamin; Klebaner, Arkadiy [Fermilab, P.O. Box 500, Batavia, IL 60510 (United States); Quack, Hans [Technische Universitaet Dresden, Dresden (Germany)

    2014-01-29T23:59:59.000Z

    Fermi National Accelerator Laboratory (FNAL) is in the process of re-commissioning a vintage CTI-4000 liquid helium coldbox, initially supplied by CTI-Cryogenics/Sulzer to Los Alamos in 1979. The coldbox was originally designed as a liquid helium refrigerator with capacity of ?1200 W at nominal 4-K. The process utilized LN{sub 2} precooling, in-series operation of two centrifugal gas bearing turboexpanders and final Joule-Thomson (J-T) expansion. At FNAL, the coldbox will be utilized as a liquefier to support 2-K operations. A process model was developed to aid in the upgrade decisions and used to determine the nominal capacity of the liquefier. Capacity upgrades are achieved by safely utilizing the internal LN2 precooler, the addition of a 3-inch reciprocating wet expansion engine and increasing the overall process pressure by recertifying two limiting pressure vessels to a higher MAWP.

  7. Chemi-microbial processing of waste tire rubber: A project overview

    SciTech Connect (OSTI)

    Romine, R.A.; Snowden-Swan, L.

    1993-12-01T23:59:59.000Z

    PNL is developing a method to use thiophillic microorganisms to devulcanize (biodesulfurize) the surface of ground rubber particles, which will improve the bonding and adhesion of the ground tire rubber into the virgin tire rubber matrix. The Chemi-microbial processing approach, introduced in this paper, is targeted at alleviating the waste tire problem in an environmentally conscious manner; it may also be applied to improve asphaltic materials and rubber and polymeric wastes to facilite their recycling. This paper outlines the logic and technical methods that will be used.

  8. Bagless transfer process and apparatus for radioactive waste confinement

    DOE Patents [OSTI]

    Maxwell, D.N.; Hones, R.H.; Rogers, M.L.

    1998-04-14T23:59:59.000Z

    A process and apparatus are provided for removing radioactive material from a glovebox, placing the material in a stainless steel storage vessel in communication with the glovebox, and sealing the vessel with a welded plug. The vessel is then severed along the weld, a lower half of the plug forming a closure for the vessel. The remaining welded plug half provides a seal for the remnant portion of the vessel and thereby maintains the sealed integrity of the glovebox. 7 figs.

  9. Bagless transfer process and apparatus for radioactive waste confinement

    DOE Patents [OSTI]

    Maxwell, David N. (Aiken, SC); Hones, Robert H. (Evans, GA); Rogers, M. Lane (Aiken, SC)

    1998-01-01T23:59:59.000Z

    A process and apparatus is provided for removing radioactive material from a glovebox, placing the material in a stainless steel storage vessel in communication with the glovebox, and sealing the vessel with a welded plug. The vessel is then severed along the weld, a lower half of the plug forming a closure for the vessel. The remaining welded plug half provides a seal for the remnant portion of the vessel and thereby maintains the sealed integrity of the glovebox.

  10. Integrated DWPF Melter System (IDMS) campaign report: Hanford Waste Vitrification Plan (HWVP) process demonstration

    SciTech Connect (OSTI)

    Hutson, N.D.

    1992-08-10T23:59:59.000Z

    Vitrification facilities are being developed worldwide to convert high-level nuclear waste to a durable glass form for permanent disposal. Facilities in the United States include the Department of Energy`s Defense Waste Processing Facility (DWPF) at the Savannah River Site, the Hanford Waste Vitrification Plant (HWVP) at the Hanford Site and the West Valley Demonstration Project (WVDP) at West Valley, NY. At each of these sites, highly radioactive defense waste will be vitrified to a stable borosilicate glass. The DWPF and WVDP are near physical completion while the HWVP is in the design phase. The Integrated DWPF Melter System (IDMS) is a vitrification test facility at the Savannah River Technology Center (SRTC). It was designed and constructed to provide an engineering-scale representation of the DWPF melter and its associated feed preparation and off-gas treatment systems. Because of the similarities of the DWPF and HWVP processes, the IDMS facility has also been used to characterize the processing behavior of a reference NCAW simulant. The demonstration was undertaken specifically to determine material balances, to characterize the evolution of offgas products (especially hydrogen), to determine the effects of noble metals, and to obtain general HWVP design data. The campaign was conducted from November, 1991 to February, 1992.

  11. Development of the fluidized bed thermal treatment process for treating mixed waste

    SciTech Connect (OSTI)

    Semones, G.B.; Williams, P.M.; Stiefvater, S.P.; Mitchell, D.L.; Roecker, B.D.

    1993-05-01T23:59:59.000Z

    A fluidized bed system is being developed at Rocky Flats for the treatment of mixed waste (a mixture of radioactive and chemically hazardous waste). The current program builds on experience gained in the 1970`s and 1980`s in tests with bench-scale, pilot-scale, and demonstration-scale fluidized bed systems. The system operates at low temperatures ({approx} 525--600{degree}C) which eliminates many of the disadvantages associated with high temperature thermal treatment processes. The process has shown the ability to destroy polychlorinated biphenyls (PCB`s) with 99.9999% (``six-nines``) destruction efficiency in tests monitored by the Environmental Protection Agency (EPA). The bed makes use of in situ neutralization of acidic off-gases by incorporating sodium carbonate (Na{sub 2}CO{sub 3}) in the bed media. This eliminates using wet scrubbers to treat the off-gas; these produce a high volume of secondary waste. Once in operation, it is expected that the fluidized bed process will yield up to a 40:1 reduction in the volume of the waste.

  12. 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.

    2002-01-02T23:59:59.000Z

    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.

  13. 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

    2002-01-01T23:59:59.000Z

    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.

  14. Liquid phase methanol reactor staging process for the production of methanol

    DOE Patents [OSTI]

    Bonnell, Leo W. (Macungie, PA); Perka, Alan T. (Macungie, PA); Roberts, George W. (Emmaus, PA)

    1988-01-01T23:59:59.000Z

    The present invention is a process for the production of methanol from a syngas feed containing carbon monoxide, carbon dioxide and hydrogen. Basically, the process is the combination of two liquid phase methanol reactors into a staging process, such that each reactor is operated to favor a particular reaction mechanism. In the first reactor, the operation is controlled to favor the hydrogenation of carbon monoxide, and in the second reactor, the operation is controlled so as to favor the hydrogenation of carbon dioxide. This staging process results in substantial increases in methanol yield.

  15. Methods of producing alkylated hydrocarbons from an in situ heat treatment process liquid

    DOE Patents [OSTI]

    Roes, Augustinus Wilhelmus Maria (Houston, TX); Mo, Weijian (Sugar Land, TX); Muylle, Michel Serge Marie (Houston, TX); Mandema, Remco Hugo (Houston, TX); Nair, Vijay (Katy, TX)

    2009-09-01T23:59:59.000Z

    A method for producing alkylated hydrocarbons is disclosed. Formation fluid is produced from a subsurface in situ heat treatment process. The formation fluid is separated to produce a liquid stream and a first gas stream. The first gas stream includes olefins. The liquid stream is fractionated to produce at least a second gas stream including hydrocarbons having a carbon number of at least 3. The first gas stream and the second gas stream are introduced into an alkylation unit to produce alkylated hydrocarbons. At least a portion of the olefins in the first gas stream enhance alkylation.

  16. Method and apparatus for measuring coupled flow, transport, and reaction processes under liquid unsaturated flow conditions

    DOE Patents [OSTI]

    McGrail, Bernard P. (Pasco, WA); Martin, Paul F. (Richland, WA); Lindenmeier, Clark W. (Richland, WA)

    1999-01-01T23:59:59.000Z

    The present invention is a method and apparatus for measuring coupled flow, transport and reaction processes under liquid unsaturated flow conditions. The method and apparatus of the present invention permit distinguishing individual precipitation events and their effect on dissolution behavior isolated to the specific event. The present invention is especially useful for dynamically measuring hydraulic parameters when a chemical reaction occurs between a particulate material and either liquid or gas (e.g. air) or both, causing precipitation that changes the pore structure of the test material.

  17. Modelling of post-fragmentation waste stream processing within UK shredder facilities

    SciTech Connect (OSTI)

    Coates, Gareth [Centre for Sustainable Manufacturing and Reuse/Recycling Technologies (SMART), Wolfson School of Mechanical and Manufacturing Engineering, Loughborough University, Loughborough LE11 3TU (United Kingdom)], E-mail: G.Coates@lboro.ac.uk; Rahimifard, Shahin [Centre for Sustainable Manufacturing and Reuse/Recycling Technologies (SMART), Wolfson School of Mechanical and Manufacturing Engineering, Loughborough University, Loughborough LE11 3TU (United Kingdom)

    2009-01-15T23:59:59.000Z

    With the introduction of producer responsibility legislation within the UK (i.e., waste electrical and electronic equipment directive and end-of-life vehicles directive), specific recycling and recovery targets have been imposed to improve the sustainability of end-of-life products. With the introduction of these targets, and the increased investment in post-fragmentation facilities, automated material separation technologies are playing an integral role within the UK's end-of-life waste management strategy. Post-fragmentation facilities utilise a range of purification technologies that target certain material attributes (e.g., density, magnetism, volume) to isolate materials from the shredded waste stream. High ferrous prices have historically meant that UK facilities have been primarily interested in recovering iron and steel, establishing processing routes that are very effective at removing these material types, but as a consequence are extremely rigid and inflexible. With the proliferation of more exotic materials within end-of-life products, combined with more stringent recycling targets, there is therefore a need to optimise the current waste reclamation processes to better realise effort-to-value returns. This paper provides a background as to the current post-fragmentation processing adopted within the UK, and describes the development of a post-fragmentation modelling approach, capable of simulating the value-added processing that a piece of automated separation equipment can have on a fragmented waste stream. These include the modelling of the inefficiencies of the technology, the effects of material entanglement on separation, determination of typical material sizing and an appreciation for compositional value. The implementation of this approach within a software decision-support system is described, before the limitations, calibration and further validation of the approach are discussed.

  18. New Waste Calcining Facility Non-Radioactive Process Decontamination

    SciTech Connect (OSTI)

    Swenson, Michael C.

    2001-09-30T23:59:59.000Z

    This report documents the results of a test of the New Calcining Facility (NWCF) process decontamination system. The decontamination system test occurred in December 1981, during non-radioactive testing of the NWCF. The purpose of the decontamination system test was to identify equipment whose design prevented effective calcine removal and decontamination. Effective equipment decontamination was essential to reduce radiation fields for in-cell work after radioactive processing began. The decontamination system test began with a pre-decontamination inspection of the equipment. The pre- decontamination inspection documented the initial condition and cleanliness of the equipment. It provided a basis for judging the effectiveness of the decontamination. The decontamination consisted of a series of equipment flushes using nitric acid and water. A post-decontamination equipment inspection determined the effectiveness of the decontamination. The pre-decontamination and post-decontamination equipment inspections were documented with photographs. The decontamination system was effective in removing calcine from most of the NWCF equipment as evidenced by little visible calcine residue in the equipment after decontamination. The decontamination test identified four areas where the decontamination system required improvement. These included the Calciner off-gas line, Cyclone off-gas line, fluidizing air line, and the Calciner baffle plates. Physical modifications to enhance decontamination were made to those areas, resulting in an effective NWCF decontamination system.

  19. New Waste Calcining Facility Non-radioactive Process Decontamination

    SciTech Connect (OSTI)

    Swenson, Michael Clair

    2001-09-01T23:59:59.000Z

    This report documents the results of a test of the New Calcining Facility (NWCF) process decontamination system. The decontamination system test occurred in December 1981, during non-radioactive testing of the NWCF. The purpose of the decontamination system test was to identify equipment whose design prevented effective calcine removal and decontamination. Effective equipment decontamination was essential to reduce radiation fields for in-cell work after radioactive processing began. The decontamination system test began with a pre-decontamination inspection of the equipment. The pre-decontamination inspection documented the initial condition and cleanliness of the equipment. It provided a basis for judging the effectiveness of the decontamination. The decontamination consisted of a series of equipment flushes using nitric acid and water. A post-decontamination equipment inspection determined the effectiveness of the decontamination. The pre-decontamination and post-decontamination equipment inspections were documented with hotographs. The decontamination system was effective in removing calcine from most of the NWCF equipment as evidenced by little visible calcine residue in the equipment after decontamination. The decontamination test identified four areas where the decontamination system required improvement. These included the Calciner off-gas line, Cyclone off-gas line, fluidizing air line, and the Calciner baffle plates. Physical modifications to enhance decontamination were made to those areas, resulting in an effective NWCF decontamination system.

  20. Process for the encapsulation and stabilization of radioactive, hazardous and mixed wastes

    DOE Patents [OSTI]

    Colombo, Peter (Patchogue, NY); Kalb, Paul D. (Wading River, NY); Heiser, III, John H. (Bayport, NY)

    1997-11-14T23:59:59.000Z

    The present invention provides a method for encapsulating and stabilizing radioactive, hazardous and mixed wastes in a modified sulfur cement composition. The waste may be incinerator fly ash or bottom ash including radioactive contaminants, toxic metal salts and other wastes commonly found in refuse. The process may use glass fibers mixed into the composition to improve the tensile strength and a low concentration of anhydrous sodium sulfide to reduce toxic metal solubility. The present invention preferably includes a method for encapsulating radioactive, hazardous and mixed wastes by combining substantially anhydrous wastes, molten modified sulfur cement, preferably glass fibers, as well as anhydrous sodium sulfide or calcium hydroxide or sodium hydroxide in a heated double-planetary orbital mixer. The modified sulfur cement is preheated to about 135.degree..+-.5.degree. C., then the remaining substantially dry components are added and mixed to homogeneity. The homogeneous molten mixture is poured or extruded into a suitable mold. The mold is allowed to cool, while the mixture hardens, thereby immobilizing and encapsulating the contaminants present in the ash.

  1. Apparatus for the processing of solid mixed waste containing radioactive and hazardous materials

    DOE Patents [OSTI]

    Gotovchikov, V.T.; Ivanov, A.V.; Filippov, E.A.

    1999-03-16T23:59:59.000Z

    Apparatus for the continuous heating and melting of a solid mixed waste bearing radioactive and hazardous materials to form separate metallic, slag and gaseous phases for producing compact forms of the waste material to facilitate disposal includes a copper split water-cooled (cold) crucible as a reaction vessel for receiving the waste material. The waste material is heated by means of the combination of a plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter. 6 figs.

  2. Apparatus for the processing of solid mixed waste containing radioactive and hazardous materials

    DOE Patents [OSTI]

    Gotovchikov, Vitaly T. (Moscow, RU); Ivanov, Alexander V. (Moscow, RU); Filippov, Eugene A. (Moscow, RU)

    1999-03-16T23:59:59.000Z

    Apparatus for the continuous heating and melting of a solid mixed waste bearing radioactive and hazardous materials to form separate metallic, slag and gaseous phases for producing compact forms of the waste material to facilitate disposal includes a copper split water-cooled (cold) crucible as a reaction vessel for receiving the waste material. The waste material is heated by means of the combination oaf plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter.

  3. The production of chemicals from food processing wastes using a novel fermenter separator. Annual progress report, January 1993--March 1994

    SciTech Connect (OSTI)

    Dale, M.C.; Venkatesh, K.V.; Choi, H.; Salicetti-Piazza, L.; Borgos-Rubio, N.; Okos, M.R.; Wankat, P.C.

    1994-03-15T23:59:59.000Z

    The basic objective of this project is to convert waste streams from the food processing industry to usable fuels and chemicals using novel bioreactors. These bioreactors should allow economical utilization of waste (whey, waste sugars, waste starch, bottling wastes, candy wastes, molasses, and cellulosic wastes) by the production of ethanol, acetone/butanol, organic acids (acetic, lactic, and gluconic), yeast diacetyl flavor, and antifungal compounds. Continuous processes incorporating various processing improvements such as simultaneous product separation and immobilized cells are being developed to allow commercial scale utilization of waste stream. The production of ethanol by a continuous reactor-separator is the process closest to commercialization with a 7,500 liter pilot plant presently sited at an Iowa site to convert whey lactose to ethanol. Accomplishments during 1993 include installation and start-up of a 7,500 liter ICRS for ethanol production at an industry site in Iowa; Donation and installation of a 200 liter yeast pilot Plant to the project from Kenyon Enterprises; Modeling and testing of a low energy system for recovery of ethanol from vapor is using a solvent absorption/extractive distillation system; Simultaneous saccharification/fermentation of raw corn grits and starch in a stirred reactor/separator; Testing of the ability of `koji` process to ferment raw corn grits in a `no-cook` process.

  4. Modeling of batch operations in the Defense Waste Processing Facility at the Savannah River Site

    SciTech Connect (OSTI)

    Smith, F.G.

    1995-02-01T23:59:59.000Z

    A computer model is in development to provide a dynamic simulation of batch operations within the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). The DWPF will chemically treat high level waste materials from the site tank farm and vitrify the resulting slurry into a borosilicate glass for permanent disposal. The DWPF consists of three major processing areas: Salt Processing Cell (SPC), Chemical Processing Cell (CPC) and the Melt Cell. Separate models have been developed for each of these process units using the SPEEDUP{trademark} software from Aspen Technology. Except for glass production in the Melt Cell, all of the chemical operations within DWPF are batch processes. Since the SPEEDUP software is designed for dynamic modeling of continuous processes, considerable effort was required to devise batch process algorithms. This effort was successful and the models are able to simulate batch operations and the dynamic behavior of the process. In this paper, we will describe the SPC model in some detail and present preliminary results from a few simulation studies.

  5. An Ionic Liquid Reaction and Separation Process for Production of Hydroxymethylfurfural from Sugars

    SciTech Connect (OSTI)

    Liu, Wei; Zheng, Feng; Li, Joanne; Cooper, Alan R.

    2014-01-01T23:59:59.000Z

    There has been world-wide interest to making plastics out of renewable biomass feedstock for recent years. Hydroxymethylfurfural (HMF) is viewed as an attractive alternate to terephthalic acid (TPA) for production of polyesters (PET) and polyamides. Conversion of sugars into HMF has been studied in numerous publications. In this work, a complete ionic liquid reaction and separation process is presented for nearly stoichiometric conversion of fructose into HMF. Different adsorbent materials are evaluated and silicalite material is demonstrated effective for isolation of 99% pure HMF from actual ionic liquid reaction mixtures and for recovery of the un-converted sugars and reaction intermediate along with the ionic liquid. Membrane-coated silicalite particles are prepared and studied for a practical adsorption process operated at low pressure drops but with separation performances comparable or better than the powder material. Complete conversion of fresh fructose feed into HMF in the recycled ionic liquid is shown under suitable reaction conditions. Stability of HMF product is characterized. A simplified process flow diagram is proposed based on these research results, and the key equipment such as reactor and adsorbent bed is sized for a plant of 200,000 ton/year of fructose processing capacity. The proposed HMF production process is much simpler than the current paraxylene (PX) manufacturing process from petroleum oil, which suggests substantial reduction to the capital cost and energy consumption be possible. At the equivalent value to PX on the molar basis, there can be a large gross margin for HMF production from fructose and/or sugars.

  6. Structural Integrity Program for the 300,000-Gallon Radioactive Liquid Waste Storage Tanks at the Idaho Nuclear Technology and Engineering Center

    SciTech Connect (OSTI)

    Bryant, J.W.; Nenni, J.A.; Yoder, T.S.

    2003-04-22T23:59:59.000Z

    This report provides a record of the Structural Integrity Program for the 300,000-gal liquid waste storage tanks and associated equipment at the Idaho Nuclear Technology and Engineering Center, as required by U.S. Department of Energy M 435.1-1, ''Radioactive Waste Management Manual.'' This equipment is known collectively as the Tank Farm Facility. The conclusion of this report is that the Tank Farm Facility tanks, vaults, and transfer systems that remain in service for storage are structurally adequate, and are expected to remain structurally adequate over the remainder of their planned service life through 2012. Recommendations are provided for continued monitoring of the Tank Farm Facility.

  7. Structural Integrity Program for the 300,000-Gallon Radioactive Liquid Waste Tanks at the Idaho Nuclear Technology and Engineering Center

    SciTech Connect (OSTI)

    Bryant, Jeffrey Whealdon; Nenni, Joseph A; Timothy S. Yoder

    2003-04-01T23:59:59.000Z

    This report provides a record of the Structural Integrity Program for the 300,000-gal liquid waste storage tanks and associated equipment at the Idaho Nuclear Technology and Engineering Center, as required by U.S. Department of Energy M 435.1-1, “Radioactive Waste Management Manual.” This equipment is known collectively as the Tank Farm Facility. The conclusion of this report is that the Tank Farm Facility tanks, vaults, and transfer systems that remain in service for storage are structurally adequate, and are expected to remain structurally adequate over the remainder of their planned service life through 2012. Recommendations are provided for continued monitoring of the Tank Farm Facility.

  8. Waste receiving and processing facility module 1, detailed design report

    SciTech Connect (OSTI)

    Not Available

    1993-10-01T23:59:59.000Z

    WRAP 1 baseline documents which guided the technical development of the Title design included: (a) A/E Statement of Work (SOW) Revision 4C: This DOE-RL contractual document specified the workscope, deliverables, schedule, method of performance and reference criteria for the Title design preparation. (b) Functional Design Criteria (FDC) Revision 1: This DOE-RL technical criteria document specified the overall operational criteria for the facility. The document was a Revision 0 at the beginning of the design and advanced to Revision 1 during the tenure of the Title design. (c) Supplemental Design Requirements Document (SDRD) Revision 3: This baseline criteria document prepared by WHC for DOE-RL augments the FDC by providing further definition of the process, operational safety, and facility requirements to the A/E for guidance in preparing the design. The document was at a very preliminary stage at the onset of Title design and was revised in concert with the results of the engineering studies that were performed to resolve the numerous technical issues that the project faced when Title I was initiated, as well as, by requirements established during the course of the Title II design.

  9. Chemistry of proposed calcination/dissolution processing of Hanford Site tank wastes

    SciTech Connect (OSTI)

    Delegard, C.H.

    1995-01-01T23:59:59.000Z

    Plans exist to separate radioactive waste stored in underground tanks at the US Department of Energy`s Hanford Site in south central Washington State into low-level and high-level fractions, and to immobilize the separate fractions in high-integrity vitrified forms for long-term disposal. Calcination with water dissolution has been proposed as a possible treatment for achieving low/high-level separation. Chemistry development activities conducted since 1992 with simulated and genuine tank waste show that calcination/dissolution destroys organic carbon and converts nitrate and nitrite to hydroxide and benign offgases. The process also dissolves significant quantities of bulk chemicals (aluminum, chromium, and phosphate), allowing their redistribution from the high-level to the low-level fraction. Present studies of the chemistry of calcination/dissolution processing of genuine wastes, conducted in the period October 1993 to September 1994, show the importance of sodium fluoride phosphate double salt in controlling phosphate dissolution. Peptization of waste solids is of concern if extensive washing occurs. Strongly oxidizing conditions imposed by calcination reactions were found to convert transition metals to soluble anions in the order chromate > manganate > > ferrate. In analogy with manganese behavior, plutonium dissolution, presumably by oxidation to more soluble anionic species, also occurs by calcination/dissolution. Methods to remove plutonium from the product low-level solution stream must be developed.

  10. UNITED STATES DEPARTMENT OF ENERGY WASTE PROCESSING ANNUAL TECHNOLOGY DEVELOPMENT REPORT 2007

    SciTech Connect (OSTI)

    Bush, S

    2008-08-12T23:59:59.000Z

    The Office of Environmental Management's (EM) Roadmap, U.S. Department of Energy--Office of Environmental Management Engineering & Technology Roadmap (Roadmap), defines the Department's intent to reduce the technical risk and uncertainty in its cleanup programs. The unique nature of many of the remaining facilities will require a strong and responsive engineering and technology program to improve worker and public safety, and reduce costs and environmental impacts while completing the cleanup program. The technical risks and uncertainties associated with cleanup program were identified through: (1) project risk assessments, (2) programmatic external technical reviews and technology readiness assessments, and (3) direct site input. In order to address these needs, the technical risks and uncertainties were compiled and divided into the program areas of: Waste Processing, Groundwater and Soil Remediation, and Deactivation and Decommissioning (D&D). Strategic initiatives were then developed within each program area to address the technical risks and uncertainties in that program area. These strategic initiatives were subsequently incorporated into the Roadmap, where they form the strategic framework of the EM Engineering & Technology Program. The EM-21 Multi-Year Program Plan (MYPP) supports the goals and objectives of the Roadmap by providing direction for technology enhancement, development, and demonstrations that will lead to a reduction of technical uncertainties in EM waste processing activities. The current MYPP summarizes the strategic initiatives and the scope of the activities within each initiative that are proposed for the next five years (FY2008-2012) to improve safety and reduce costs and environmental impacts associated with waste processing; authorized budget levels will impact how much of the scope of activities can be executed, on a year-to-year basis. As a result of the importance of reducing technical risk and uncertainty in the EM Waste Processing programs, EM-21 has focused considerable effort on identifying the key areas of risk in the Waste Processing programs. The resulting summary of technical risks and needs was captured in the Roadmap. The Roadmap identifies key Waste Processing initiative areas where technology development work should be focused. These areas are listed below, along with the Work Breakdown Structure (WBS) designation given to each initiative area. The WBS designations will be used throughout this document.

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

    SciTech Connect (OSTI)

    Ramsey, William Gene

    2013-08-15T23:59:59.000Z

    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.

  12. HUMAN MACHINE INTERFACE (HMI) EVALUATION OF ROOMS TA-50-1-60/60A AT THE RADIOACTIVE LIQUID WASTE TREATMENT FACILITY (RLWTF)

    SciTech Connect (OSTI)

    Gilmore, Walter E. [Los Alamos National Laboratory; Stender, Kerith K. [Los Alamos National Laboratory

    2012-08-29T23:59:59.000Z

    This effort addressed an evaluation of human machine interfaces (HMIs) in Room TA-50-1-60/60A of the Radioactive Liquid Waste Treatment Facility (RLWTF). The evaluation was performed in accordance with guidance outlined in DOE-STD-3009, DOE Standard Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, 2006 [DOE 2006]. Specifically, Chapter 13 of DOE 2006 highlights the 10 CFR 830, Nuclear Safety Management, 2012, [CFR 2012] and DOE G 421.1-2 [DOE 2001a] requirements as they relate to the human factors process and, in this case, the safety of the RLWTF. The RLWTF is a Hazard Category 3 facility and, consequently, does not have safety-class (SSCs). However, safety-significant SSCs are identified. The transuranic (TRU) wastewater tanks and associated piping are the only safety-significant SSCs in Rooms TA-50-1-60/60A [LANL 2010]. Hence, the human factors evaluation described herein is only applicable to this particular assemblage of tanks and piping.

  13. Catalytic co-processing of coal with bitumen and bitumen derived liquids

    SciTech Connect (OSTI)

    Chakma, A.; Zaman, J. [Univ. of Calgary, Alberta (Canada)

    1993-12-31T23:59:59.000Z

    Experimental studies on the co-processing of coal with bitumen and bitumen derived liquids are described. A subbituminous coal was coprocessed with Athabasca bitumen and its various liquid fractions in a batch autoclave under hydrogen pressure at reaction temperatures varying from 400 to 440{degrees}C. Both thermal and catalytic coprocessing experiments were conducted. The catalysts used were molten halide type and included ZnCl{sub 2}, MoCl{sub 5}, KCl, CuCl, and SnCl{sub 2}. Higher reaction temperature resulted in higher conversion of asphaltenes into both maltenes and coke and gases. As a result the H/C atomic ratio of the unconverted asphaltenes decreased with temperature. Higher reaction time on the other hand allowed maltenes to be converted to asphaltenes. While all the catalysts tested had catalytic effects on asphaltene conversion, MoCl{sub 5} was found to provide the highest conversion of asphaltenes due to its ability to hydrogenate the radicals formed due to asphaltene cracking. Processing of coal with bitumen derived liquids provided higher yields than those obtained with virgin bitumen. The H/C ratios were also higher for the products obtained with bitumen derived liquids.

  14. 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.

    1986-09-01T23:59:59.000Z

    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.

  15. The physical separation and recovery of metals from wastes. Process engineering for the chemical, metals and minerals industries, Volume 1

    SciTech Connect (OSTI)

    Veasey, T.J.; Wilson, R.J. (eds.) (Univ. of Birmingham (United Kingdom). School of Chemical Engineering); Squires, D.M. (ed.) (Newell Engineering Ltd., Redditch (United Kingdom))

    1993-01-01T23:59:59.000Z

    This book deals with the physical processes used for the separation of secondary metals from waste sources. The introduction briefly considers the history of the secondary metals industries, defines the terms used in materials recycling and discusses the potential for resource recovery and improved processing. A comprehensive survey is given of the unit operations employed for metals recovery and reclamation, and this is followed by detailed descriptions of processes used to treat fragmentized metal wastes and granulated metal wastes. The final chapter reviews the processing of urban wastes for metals recovery, and gives details of modern plant and practices. The volume aims to bring together technical information on metals recovery from a wide range of sources in order to give a unified review of an important engineering and environmental topic. Topics include: general definitions used in materials recycling; the potential for resource recovery; secondary metals; ranking of scrap; the potential for improved processing; comminution; physical separation methods; the scrap industry; automobile composition; shredders; non-magnetic processing; metal reclamation processes; waste tire processing; battery processing; thermal processing systems; composition of urban waste; and material recovery.

  16. Commercial-Scale Demonstration of the Liquid Phase Methanol (LPMEOH(TM)) Process

    SciTech Connect (OSTI)

    None

    1996-12-31T23:59:59.000Z

    The Liquid Phase Methanol (LPMEOH(TM)) Demonstration Project at Kingsport, Tennessee, is a $213.7 million cooperative agreement between the U.S. Department of Energy (DOE) and Air Products Liquid Phase Conversion Company, L.P. (the Partnership). The LPMEOIWM Process Demonstration Unit is being built at a site located at the Eastman Chemical Company (Eastman) complex in Kingsport. During this quarter, the Cooperative Agreement was modified (Mod AO11) on 8 October 1996, authorizing the transition born Budget Period No. 2 (Design and Construction) to the . final Budget Period (Commissioning, Start-up, and Operation), A draft Topical Report on Process Economics Studies concludes that methanol coproduction with integrated gasification combined cycle (IGCC) electric power utilizing the LPMEOW process technology, will be competitive in serving local market needs. Planning for a proof-of- concept test run of the liquid phase dimethyl ether (DME) process at the LaPorte Alternative Fuels Development Unit (AFDU) was recommended; and a deeision to proceed is pending. Construction (Task 2.2) is 97'Mo complete, asof31 December 1996. Completion of pipe pressure testing has taken longer than expected. This will delay completion of construction by about three weeks. Commissioning activities (Task 2.3) commenced in mid-October of 1996, and the demonstration unit is scheduled to be mechanically complete on 24 January 1997.

  17. High Level Waste Tank Farm Replacement Project for the Idaho Chemical Processing Plant at the Idaho National Engineering Laboratory. Environmental Assessment

    SciTech Connect (OSTI)

    Not Available

    1993-06-01T23:59:59.000Z

    The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0831, for the construction and operation of the High-Level Waste Tank Farm Replacement (HLWTFR) Project for the Idaho Chemical Processing Plant located at the Idaho National Engineering Laboratory (INEL). The HLWTFR Project as originally proposed by the DOE and as analyzed in this EA included: (1) replacement of five high-level liquid waste storage tanks with four new tanks and (2) the upgrading of existing tank relief piping and high-level liquid waste transfer systems. As a result of the April 1992 decision to discontinue the reprocessing of spent nuclear fuel at INEL, DOE believes that it is unlikely that the tank replacement aspect of the project will be needed in the near term. Therefore, DOE is not proposing to proceed with the replacement of the tanks as described in this-EA. The DOE`s instant decision involves only the proposed upgrades aspect of the project described in this EA. The upgrades are needed to comply with Resource Conservation and Recovery Act, the Idaho Hazardous Waste Management Act requirements, and the Department`s obligations pursuant to the Federal Facilities Compliance Agreement and Consent Order among the Environmental Protection Agency, DOE, and the State of Idaho. The environmental impacts of the proposed upgrades are adequately covered and are bounded by the analysis in this EA. If DOE later proposes to proceed with the tank replacement aspect of the project as described in the EA or as modified, it will undertake appropriate further review pursuant to the National Environmental Policy Act.

  18. Sampling, preservation, and analytical methods research plan - liquid redox sulfur recovery technologies: Stretford process. Topical report

    SciTech Connect (OSTI)

    Trofe, T.W.

    1986-11-01T23:59:59.000Z

    GRI has developed a sampling, preservation, and analytical (SPandA) methods research plan for developing and validating analytical methodologies for liquid redox sulfur recovery processes (e.g., Stretford process). The document describes the technical approach which will be used to direct research activities to develop SPandA methodologies to analyze gaseous, aqueous, and solid process streams from the Stretford sulfur recovery process. The primary emphasis is on developing and validating methodologies for analyzing vanadium (IV) and vanadium (V), anthraquinone disulphonic acids (ADA), polysulfide-sulfur, sulfide-sulfur, thiosulfate, sulfate, thiocyanate, total soluble sulfur, alkalinity, pH, total dissolved solids, total suspended solids, and dissolved oxygen in aqueous process streams. The document includes descriptions of the process streams and chemical species, selection of candidate analytical methods, and technical approach for methods development and validation.

  19. Process Options Description for Vitrification Flowsheet Model of INEEL Sodium Bearing Waste

    SciTech Connect (OSTI)

    Nichols, Todd Travis; Taylor, Dean Dalton; Lauerhass, Lance; Barnes, Charles Marshall

    2001-02-01T23:59:59.000Z

    The purpose of this document is to provide the technical information to Savannah River Site (SRS) personnel that is required for the development of a basic steady-state process simulation of the vitrification treatment train of sodium bearing waste (SBW) at Idaho National Engineering and nvironmental Laboratory (INEEL). INEEL considers simulation to have an important role in the integration/optimization of treatment process trains for the High Level Waste (HLW) Program. This project involves a joint Technical Task Plan (TTP ID77WT31, Subtask C) between SRS and INEEL. The work scope of simulation is different at the two sites. This document addresses only the treatment of SBW at INEEL. The simulation model(s) is to be built by SRS for INEEL in FY-2001.

  20. High liquid yield process for retorting various organic materials including oil shale

    DOE Patents [OSTI]

    Coburn, Thomas T. (Livermore, CA)

    1990-01-01T23:59:59.000Z

    This invention is a continuous retorting process for various high molecular weight organic materials, including oil shale, that yields an enhanced output of liquid product. The organic material, mineral matter, and an acidic catalyst, that appreciably adsorbs alkenes on surface sites at prescribed temperatures, are mixed and introduced into a pyrolyzer. A circulating stream of olefin enriched pyrolysis gas is continuously swept through the organic material and catalyst, whereupon, as the result of pyrolysis, the enhanced liquid product output is provided. Mixed spent organic material, mineral matter, and cool catalyst are continuously withdrawn from the pyrolyzer. Combustion of the spent organic material and mineral matter serves to reheat the catalyst. Olefin depleted pyrolysis gas, from the pyrolyzer, is enriched in olefins and recycled into the pyrolyzer. The reheated acidic catalyst is separated from the mineral matter and again mixed with fresh organic material, to maintain the continuously cyclic process.

  1. A high liquid yield process for retorting various organic materials including oil shale

    DOE Patents [OSTI]

    Coburn, T.T.

    1988-07-26T23:59:59.000Z

    This invention is a continuous retorting process for various high molecular weight organic materials, including oil shale, that yields an enhanced output of liquid product. The organic material, mineral matter, and an acidic catalyst, that appreciably adsorbs alkenes on surface sites at prescribed temperatures, are mixed and introduced into a pyrolyzer. A circulating stream of olefin enriched pyrolysis gas is continuously swept through the organic material and catalyst, whereupon, as the result of pyrolysis, the enhanced liquid product output is provided. Mixed spent organic material, mineral matter, and cool catalyst are continuously withdrawn from the pyrolyzer. Combustion of the spent organic material and mineral matter serves to reheat the catalyst. Olefin depleted pyrolysis gas, from the pyrolyzer, is enriched in olefins and recycled into the pyrolyzer. The reheated acidic catalyst is separated from the mineral matter and again mixed with fresh organic material, to maintain the continuously cyclic process. 2 figs.

  2. Microwave applicator for in-drum processing of radioactive waste slurry

    DOE Patents [OSTI]

    White, T.L.

    1994-06-28T23:59:59.000Z

    A microwave applicator for processing of radioactive waste slurry uses a waveguide network which splits an input microwave of TE[sub 10] rectangular mode to TE[sub 01] circular mode. A cylindrical body has four openings, each receiving 1/4 of the power input. The waveguide network includes a plurality of splitters to effect the 1/4 divisions of power. 4 figures.

  3. Electrodialysis-based separation process for salt recovery and recycling from waste water

    DOE Patents [OSTI]

    Tsai, S.P.

    1997-07-08T23:59:59.000Z

    A method for recovering salt from a process stream containing organic contaminants is provided, comprising directing the waste stream to a desalting electrodialysis unit so as to create a concentrated and purified salt permeate and an organic contaminants-containing stream, and contacting said concentrated salt permeate to a water-splitting electrodialysis unit so as to convert the salt to its corresponding base and acid. 6 figs.

  4. Electrodialysis-based separation process for salt recovery and recycling from waste water

    DOE Patents [OSTI]

    Tsai, Shih-Perng (Naperville, IL)

    1997-01-01T23:59:59.000Z

    A method for recovering salt from a process stream containing organic contaminants is provided, comprising directing the waste stream to a desalting electrodialysis unit so as to create a concentrated and purified salt permeate and an organic contaminants containing stream, and contacting said concentrated salt permeate to a water-splitting electrodialysis unit so as to convert the salt to its corresponding base and acid.

  5. Proposed Use of a Constructed Wetland for the Treatment of Metals in the S-04 Outfall of the Defense Waste Processing Facility at the Savannah River Site

    SciTech Connect (OSTI)

    Glover, T.

    1999-11-23T23:59:59.000Z

    The DWPF is part of an integrated waste treatment system at the SRS to treat wastes containing radioactive contaminants. In the early 1980s the DOE recognized that there would be significant safety and cost advantages associated with immobilizing the radioactive waste in a stable solid form. The Defense Waste Processing Facility was designed and constructed to accomplish this task.

  6. Tank 241-CX-70 waste removal and packaging

    SciTech Connect (OSTI)

    DuVon, D.K.

    1993-06-01T23:59:59.000Z

    Tank 241-CX-70, located on the Hanford Site in Washington State, is a 30,000 gal single-shell storage tank built in 1952 to hold high-level process waste from pilot tests of the reduction-oxidation process. In 1979 decommissioning operations were begun by pumping liquid waste from the tank to the double-shell tank (DST) 101-AY. Not all the waste was removed at that time. Approximately 10,300 gal of sludge remained. On September 25, 1987, operations were resumed to remove the remaining waste using a sluicing and pumping method. This report documents the final removal of waste from Tank 241-CX-70.

  7. Tank 241-CX-70 waste removal and packaging

    SciTech Connect (OSTI)

    DuVon, D.K.

    1993-01-01T23:59:59.000Z

    Tank 241-CX-70, located on the Hanford Site in Washington State, is a 30,000 gal single-shell storage tank built in 1952 to hold high-level process waste from pilot tests of the reduction-oxidation process. In 1979 decommissioning operations were begun by pumping liquid waste from the tank to the double-shell tank (DST) 101-AY. Not all the waste was removed at that time. Approximately 10,300 gal of sludge remained. On September 25, 1987, operations were resumed to remove the remaining waste using a sluicing and pumping method. This report documents the final removal of waste from Tank 241-CX-70.

  8. Adsorption of Ruthenium, Rhodium and Palladium from Simulated High-Level Liquid Waste by Highly Functional Xerogel - 13286

    SciTech Connect (OSTI)

    Onishi, Takashi [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan)] [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan); Koyama, Shin-ichi [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan)] [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan); Mimura, Hitoshi [Dept. of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2,Aoba-ku, Sendai-shi, Miyagi-ken, 980-8579 (Japan)] [Dept. of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2,Aoba-ku, Sendai-shi, Miyagi-ken, 980-8579 (Japan)

    2013-07-01T23:59:59.000Z

    Fission products are generated by fission reactions in nuclear fuel. Platinum group (Pt-G) elements, such as palladium (Pd), rhodium (Rh) and ruthenium (Ru), are also produced. Generally, Pt-G elements play important roles in chemical and electrical industries. Highly functional xerogels have been developed for recovery of these useful Pt-G elements from high - level radioactive liquid waste (HLLW). An adsorption experiment from simulated HLLW was done by the column method to study the selective adsorption of Pt-G elements, and it was found that not only Pd, Rh and Ru, but also nickel, zirconium and tellurium were adsorbed. All other elements were not adsorbed. Adsorbed Pd was recovered by washing the xerogel-packed column with thiourea solution and thiourea - nitric acid mixed solution in an elution experiment. Thiourea can be a poison for automotive exhaust emission system catalysts, so it is necessary to consider its removal. Thermal decomposition and an acid digestion treatment were conducted to remove sulfur in the recovered Pd fraction. The relative content of sulfur to Pd was decreased from 858 to 0.02 after the treatment. These results will contribute to design of the Pt-G element separation system. (authors)

  9. Development of low level liquid waste treatment systems: April-September 1981

    SciTech Connect (OSTI)

    Williams, M.K.; Colvin, C.M.; Bond, W.H.

    1982-03-05T23:59:59.000Z

    The pilot plant reverse osmosis system was demonstrated to be effective in removing large percentages of cobalt-60, iodine-125, and a mixture of cesium-137, cobalt-60, and iodine-125 from two types of aqueous streams. The effectiveness of three membrane porosities, 0, 50, and 97% salt rejection, were explored with each isotope. The 97% salt rejection membrane was the most effective in each experiment. Removals as high as 97.5% of the cobalt, 92.9% of the iodine and 95.1% of the combined isotopes were achieved. The effect of possibly interfering factors on the adsorbence of cobalt-60 and iodine-129 on selected ion exchange resins were investigated. The factors thought to affect cobalt-60 adsorption were (OH/sup -/), (NH/sub 4//sup +/), and (SO/sub 3//sup =/). None of the seven factors investigated had any effect on iodine-129 adsorption. Cesium-137 was removed from a 4600-gal aqueous waste containing a large amount of sodium hydroxide by treatment with sodium tetraphenyl boron. The cesium concentration of the supernatant portion was reduced from 570 to 4 counts/min/ml.

  10. Commercial-Scale Demonstration of the Liquid Phase Methanol (LOMEOH(TM)) Process

    SciTech Connect (OSTI)

    None

    1996-03-31T23:59:59.000Z

    The Liquid Phase Methanol (LPMEOEP") Demonstration Project at K.ingsport, Tennessee, is a $213.7 million cooperative agreement between the U.S. Department of Energy (DOE) and Air Products Liquid Phase Conversion Company, L, P. (the Partnership). The LPMEOHY Process Demonstration Unit is being built at a site located at the Eastman Chemical Company (Eastman) complex in Kingsport. On 4 October 1994, Air Products and Chemicals, Inc. (Air Products) and signed the agreements that would form the Partnership, secure the demonstration site, and provide the financial commitment and overall project management for the project. These partnership agreements became effective on 15 March 1995, when DOE authorized the commencement of Budget Period No. 2 (Mod. AO08 to the Cooperative Agreement). The Partnership has subcontracted with Air Products to provide the overall management of the project, and to act as the primary interface with DOE. As subcontractor to the Partnership, Air Products will also provide the engineering design, procurement, construction, and commissioning of the LPMEOHTM Process Demonstration Unit, and will provide the technical and engineering supervision needed to conduct the operational testing program required as part of the project. As subcontractor to Air Products, Eastman will be responsible for operation of the LPMEOHTM Process Demonstration Unit, and for the interconnection and supply of synthesis gas, utilities, product storage, and other needed sewices. The project involves the construction of an 80,000 gallons per day (260 tons-per-day (TPD)) methanol unit utilizing coal-derived synthesis gas fi-om Eastman's integrated coal gasification facility. The new equipment consists of synthesis gas feed preparation and compression facilities, the liquid phase reactor and auxiliaries, product distillation facilities, and utilities. The technology to be demonstrated is the product of a cooperative development effort by Air Products and DOE in a program that started in 1981. Developed to enhance electric power generation using integrated gasification combined cycle (IGCC) technology, the LPMEOHTM process is ideally suited for directly processing gases produced by modern day coal gasifiers. Originally tested at a small 3,200 gallons per day, DOE-owned experimental unit in LaPorte, Texas, the technology provides several improvements essential for the economic coproduction of methanol and electricity directly from gasified coal. This liquid phase process suspends fine catalyst particles in an inert liquid, forming a slurry. The slurry dissipates the heat of the chemical reaction away from the catalyst surface, protecting the catalyst and allowing the methanol synthesis reaction to proceed at higher rates.

  11. Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering

    DOE Patents [OSTI]

    Vijayan, Sivaraman (Deep River, CA); Wong, Chi F. (Pembroke, CA); Buckley, Leo P. (Deep River, CA)

    1994-01-01T23:59:59.000Z

    In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved.

  12. Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering

    DOE Patents [OSTI]

    Vijayan, S.; Wong, C.F.; Buckley, L.P.

    1994-11-22T23:59:59.000Z

    In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved. 1 fig.

  13. High-level waste at Hanford: Potential for waste loading maximization

    SciTech Connect (OSTI)

    Hrma, P.; Bailey, A.W. [Pacific Northwest Lab., Richland, WA (United States)

    1995-12-31T23:59:59.000Z

    The loading of Hanford nuclear waste in borosilicate glass is limited by phase-related phenomena, such as crystallization or formation of immiscible liquids, and by the breakdown of the glass structure due to an excessive concentration of modifiers. The phase-related phenomena cause both processing and product quality problems. The deterioration of the product durability determines the ultimate waste loading limit if all processing problems are resolved. Concrete examples and mass-balance based calculations show that a substantial potential exists for increasing waste loading of high-level wastes that contain a large fraction of refractory components.

  14. Liquid fuels from co-processing coal with bitumen or heavy oil: A review

    SciTech Connect (OSTI)

    Moschopedis, S.E.; Hepler, L.G.

    1987-01-01T23:59:59.000Z

    Coal, bitumen and heavy oil (and various pitches, resids, etc.) are similar in that they require more substantial treatment than does conventional light oil to yield useful liquid fuels. The authors provide a brief and selective review of technologies for liquefying coal, followed by consideration of co-processing coal with bitumen/heavy oil. Such co-processing may be considered as use of bitumen/heavy oil as a solvent and/or hydrogen donor in liquefaction of coal, or as the use of coal to aid upgrading bitumen/heavy oil.

  15. Recovery of Valuable Chlorosilane Intermediates by a Novel Waste Conversion Process, Phase IIIB (Progress)

    SciTech Connect (OSTI)

    Kurt E. Anderson

    2000-03-31T23:59:59.000Z

    From June 1998 through September 1999, direct process residue (DPR, a waste byproduct) hydrogenolysis has been studied at a large pilot plant within Dow Corning's Carrollton, KY, facility. The system reacts filtered DPR with chlorosilane monomers at high temperature and pressure. The process routinely demonstrates DPR conversions from 59% to 89% on a monthly basis. The reaction product contains high concentrations of valuable monomers such as dimethyldichlorosilane and methyldichlorosilane. An expansion of the current unit's capacity is planned to be on-line by the end of CY2000. Furthermore, a larger DPR hydrogenolysis reactor based on these results is being designed for operation in Europe at Dow Corning's Barry, Wales, site.

  16. Chemical Engineering Division fuel cycle programs. Quarterly progress report, April-June 1979. [Pyrochemical/dry processing; waste encapsulation in metal; transport in geologic media

    SciTech Connect (OSTI)

    Steindler, M.J.; Ader, M.; Barletta, R.E.

    1980-09-01T23:59:59.000Z

    For pyrochemical and dry processing materials development included exposure to molten metal and salt of Mo-0.5% Ti-0.07% Ti-0.01% C, Mo-30% W, SiC, Si/sub 2/ON/sub 2/, ZrB/sub 2/-SiC, MgAl/sub 2/O/sub 4/, Al/sub 2/O/sub 3/, AlN, HfB/sub 2/, Y/sub 2/O/sub 3/, BeO, Si/sub 3/N/sub 4/, nickel nitrate-infiltrated W, W-coated Mo, and W-metallized alumina-yttria. Work on Th-U salt transport processing included solubility of Th in liquid Cd, defining the Cd-Th and Cd-Mg-Th phase diagrams, ThO/sub 2/ reduction experiments, and electrolysis of CaO in molten salt. Work on pyrochemical processes and associated hardware for coprocessing U and Pu in spent FBR fuels included a second-generation computer model of the transport process, turntable transport process design, work on the U-Cu-Mg system, and U and Pu distribution coefficients between molten salt and metal. Refractory metal vessels are being service-life tested. The chloride volatility processing of Th-based fuel was evaluated for its proliferation resistance, and a preliminary ternary phase diagram for the Zn-U-Pu system was computed. Material characterization and process analysis were conducted on the Exportable Pyrochemical process (Pyro-Civex process). Literature data on oxidation of fissile metals to oxides were reviewed. Work was done on chemical bases for the reprocessing of actinide oxides in molten salts. Flowsheets are being developed for the processing of fuel in molten tin. Work on encapsulation of solidified radioactive waste in metal matrix included studies of leach rate of crystalline waste materials and of the impact resistance of metal-matrix waste forms. In work on the transport properties of nuclear waste in geologic media, adsorption of Sr on oolitic limestone was studied, as well as the migration of Cs in basalt. Fitting of data on the adsorption of iodate by hematite to a mathematical model was attempted.

  17. Status of development of actinide blanket processing flowsheets for accelerator transmutation of nuclear waste

    SciTech Connect (OSTI)

    Dewey, H.J.; Jarvinen, G.D.; Marsh, S.F.; Schroeder, N.C.; Smith, B.F.; Villarreal, R.; Walker, R.B.; Yarbro, S.L.; Yates, M.A.

    1993-09-01T23:59:59.000Z

    An accelerator-driven subcritical nuclear system is briefly described that transmutes actinides and selected long-lived fission products. An application of this accelerator transmutation of nuclear waste (ATW) concept to spent fuel from a commercial nuclear power plant is presented as an example. The emphasis here is on a possible aqueous processing flowsheet to separate the actinides and selected long-lived fission products from the remaining fission products within the transmutation system. In the proposed system the actinides circulate through the thermal neutron flux as a slurry of oxide particles in heavy water in two loops with different average residence times: one loop for neptunium and plutonium and one for americium and curium. Material from the Np/Pu loop is processed with a short cooling time (5-10 days) because of the need to keep the total actinide inventory, low for this particular ATW application. The high radiation and thermal load from the irradiated material places severe constraints on the separation processes that can be used. The oxide particles are dissolved in nitric acid and a quarternary, ammonium anion exchanger is used to extract neptunium, plutonium, technetium, and palladium. After further cooling (about 90 days), the Am, Cm and higher actinides are extracted using a TALSPEAK-type process. The proposed operations were chosen because they have been successfully tested for processing high-level radioactive fuels or wastes in gram to kilogram quantities.

  18. 1M. Panahi, S. Skogestad ' Controlled Variables Selection for a Natural Gas to Liquids (GTL) process' Controlled Variables Selection for a

    E-Print Network [OSTI]

    Skogestad, Sigurd

    1M. Panahi, S. Skogestad ' Controlled Variables Selection for a Natural Gas to Liquids (GTL) process' Controlled Variables Selection for a Natural Gas to Liquids (GTL) process Mehdi Panahi Sigurd for a Natural Gas to Liquids (GTL) process' Skogestad plantwide control procedure* I Top Down · Step 1: Identify

  19. Commercial-Scale Demonstration of the Liquid Phase Methanol (LPMEOH(TM)) Process

    SciTech Connect (OSTI)

    None

    1997-09-30T23:59:59.000Z

    The Liquid Phase Methanol (LPMEOH) Demonstration Project at Kingsport, Tennessee, is a $213.7 million cooperative agreement between the U.S. Department of Energy (DOE) and Air Products Liquid Phase Conversion Company, L.P. (the Partnership). Ak Products and Chemicals, Inc. (Air Products) and Eastman Chemical Company (Eastman) formed the Partnership to execute the Demonstration Project. The LPMEOITM Process Demonstration Unit was built at a site located at the Eastman complex in Kingsport. During this reporting period, DOE accepted the recommendation to continue with dimethyl ether (DME) design verification testing (DVT). DME design verification testing studies show the liquid phase DME (LPDME) process will have a significant economic advantage for the coproduction of DME for local markets. An LPDME catalyst system with reasonable long-term activity and stzibility is being developed. Planning for a proof-of-concept test run at the LaPorte Alternative Fuels Development Unit (AFDU) was recommended. DOE issued a letter dated 31 July 1997 accepting the recommendation to continue design verification testing. In order to allow for scale-up of the manufacturing technique for the dehydration catalyst from the pilot plant to the commercial scale, the time required to produce the catalyst to the AFDU has slipped. The new estimated delivery date is 01 June 1998.

  20. Enclosure 3 DOE Response to EPA Question Regarding "High-Level Liquid Radioactive Waste"

    E-Print Network [OSTI]

    was it removed?). I don't think the removal in Appendix F of the old Appendix D language about the hulls;Enclosure 3 In the Bismuth Phosphate process, the cladding (or coating) hulls are first removed). The fuel is then dissolved into its constituent parts consisting of uranium, transuranic elements (e

  1. Waste Processing To Support {sup 99}Mo Production at Sandia National Laboratories

    SciTech Connect (OSTI)

    Longley, Susan; Carson, Susan; McDonald, Marion

    1997-06-01T23:59:59.000Z

    As part of the Isotope Production Program at Sandia National Laboratories New Mexico (SNL/NM), procedures are being finalized for the production of {sup 99}Mo from the irradiation of {sup 235}U-coated stainless steel targets at the Technical Area (TA) V reactor and hot cell facilities. Methods have been identified and tested for the management of the non-product (waste) material as the final step in the production process. These methods were developed utilizing the waste material from a series of cold and hot tests, beginning with depleted uranium powder and culminating with a test involving an irradiated {sup 235}U target with an initial fission product inventory of approximately 18,000 Ci at the end of the irradiation cycle.

  2. Standard practice for fluorescent liquid penetrant testing using the Solvent-Removable process

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    1.1 This practice covers procedures for fluorescent penetrant examination utilizing the solvent-removable process. It is a nondestructive testing method for detecting discontinuities that are open to the surface, such as cracks, seams, laps, cold shuts, laminations, isolated porosity, through leaks, or lack of fusion and is applicable to in-process, final, and maintenance examination. It can be effectively used in the examination of nonporous, metallic materials, both ferrous and nonferrous, and of nonmetallic materials such as glazed or fully densified ceramics and certain nonporous plastics and glass. 1.2 This practice also provides a reference: 1.2.1 By which a fluorescent penetrant examination solvent-removable process recommended or required by individual organizations can be reviewed to ascertain its applicability and completeness. 1.2.2 For use in the preparation of process specifications dealing with the fluorescent solvent-removable liquid penetrant examination of materials and parts. Agreement by...

  3. Analysis of tritium extraction from liquid lithium by permeation window and solid gettering processes

    SciTech Connect (OSTI)

    Takeda, T. [Japan Atomic Energy Research Inst., Ibaraki-ken (Japan); Ying, A.Y.; Abdou, M.A. [Univ. of California, Los Angeles, CA (United States)

    1994-12-31T23:59:59.000Z

    Tritium recovery from liquid lithium at low concentration is an important problem for liquid metal breeder-blanket in a fusion reactor. Previous studies have identified tritium recovery methods including molten salt extraction, gettering recovery, permeation window, and vacuum distillation. In this paper, the authors focus on the numerical studies on tritium extraction by permeation window and gettering processes. These studies include for example: dynamic tritium concentration variation along the flow direction, tritium inventory distributions in the permeator and getter bed, along with the effect of dispersion on extraction efficiency. Using a model description makes it possible to determine functional dependence and provide insight into the interrelationships of the various operating conditions and material properties which may affect the behavior of tritium in the material. Clearly, reliable material properties (such as diffusivity, solubility, etc.) are essential for realistic evaluations.

  4. Void Prediction During Liquid Composite Molding Processes: Wetting and Capillary Phenomena

    E-Print Network [OSTI]

    Abdelwahed, Mohamed Amine Ben; Bizet, Laurent; Bréard, Joël

    2012-01-01T23:59:59.000Z

    The aim of this work is to contribute in improving fibrous preforms impregnation for Liquid Composite Molding (LCM) processes. The void prediction in LCM sparks off interest within the Composite Material elaboration because it represents a significant issue to keep the expected mechanical properties of the final product. The liquid properties, the preform geometry and the flow conditions impact the void or bubble entrapped inside and outside the yarns. Nevertheless, due to the complex geometry of the reinforcement, experimental characterization of bubble formation remains delicate. Thus, our study deals with two simple model networks representing connected pores so called "Pore Doublet Model". A first is considering two capillaries converging on a node (T-junction) and a second is representing two capillaries interconnected with a supplying principle. In this paper, we emphasize on microfluidic and millifluidic approaches where wetting and capillary forces are significant during bubble formation mechanism.

  5. Waste-to-Energy Workshop Agenda

    Broader source: Energy.gov [DOE]

    The Bioenergy Technologies Office (BETO) at the Department of Energy aims to identify and address key technical barriers to the commercial deployment of liquid transportation fuels from waste feedstocks. As a part of this effort, BETO is organizing a Waste-to-Energy Roadmapping workshop. Workshop participants will join facilitated breakout sessions to discuss anaerobic digestion, hydrothermal liquefaction, and other processes that make productive use of wastewater residuals, biosolids, foodstuffs, and organic municipal solid waste. These discussions will be synthesized and used in developing a waste-to-energy technology roadmap.

  6. The National Nuclear Laboratory's Approach to Processing Mixed Wastes and Residues - 13080

    SciTech Connect (OSTI)

    Greenwood, Howard; Docrat, Tahera; Allinson, Sarah J.; Coppersthwaite, Duncan P.; Sultan, Ruqayyah; May, Sarah [National Nuclear Laboratory, Springfields, Preston, UK, PR4 0XJ (United Kingdom)] [National Nuclear Laboratory, Springfields, Preston, UK, PR4 0XJ (United Kingdom)

    2013-07-01T23:59:59.000Z

    The National Nuclear Laboratory (NNL) treats a wide variety of materials produced as by-products of the nuclear fuel cycle, mostly from uranium purification and fuel manufacture but also including materials from uranium enrichment and from the decommissioning of obsolete plants. In the context of this paper, treatment is defined as recovery of uranium or other activity from residues, the recycle of uranium to the fuel cycle or preparation for long term storage and the final disposal or discharge to the environment of the remainder of the material. NNL's systematic but flexible approach to residue assessment and treatment is described in this paper. The approach typically comprises up to five main phases. The benefits of a systematic approach to waste and residue assessments and processing are described in this paper with examples used to illustrate each phase of work. Benefits include early identification of processing routes or processing issues and the avoidance of investment in inappropriate and costly plant or processes. (authors)

  7. COMMERCIAL-SCALE DEMONSTRATION OF THE LIQUID PHASE METHANOL (LPMEOH) PROCESS

    SciTech Connect (OSTI)

    E.C. Heydorn; B.W. Diamond; R.D. Lilly

    2003-06-01T23:59:59.000Z

    This project, which was sponsored by the U.S. Department of Energy (DOE) under the Clean Coal Technology Program to demonstrate the production of methanol from coal-derived synthesis gas (syngas), has completed the 69-month operating phase of the program. The purpose of this Final Report for the ''Commercial-Scale Demonstration of the Liquid Phase Methanol (LPMEOH{trademark}) Process'' is to provide the public with details on the performance and economics of the technology. The LPMEOH{trademark} Demonstration Project was a $213.7 million cooperative agreement between the DOE and Air Products Liquid Phase Conversion Company, L.P. (the Partnership). The DOE's cost share was $92,708,370 with the remaining funds coming from the Partnership. The LPMEOH{trademark} demonstration unit is located at the Eastman Chemical Company (Eastman) chemicals-from-coal complex in Kingsport, Tennessee. The technology was the product of a cooperative development effort by Air Products and Chemicals, Inc. (Air Products) and DOE in a program that started in 1981. Developed to enhance electric power generation using integrated gasification combined cycle (IGCC) technology, the LPMEOH{trademark} Process is ideally suited for directly processing gases produced by modern coal gasifiers. Originally tested at the Alternative Fuels Development Unit (AFDU), a small, DOE-owned process development facility in LaPorte, Texas, the technology provides several improvements essential for the economic coproduction of methanol and electricity directly from gasified coal. This liquid phase process suspends fine catalyst particles in an inert liquid, forming a slurry. The slurry dissipates the heat of the chemical reaction away from the catalyst surface, protecting the catalyst, and allowing the methanol synthesis reaction to proceed at higher rates. The LPMEOH{trademark} Demonstration Project accomplished the objectives set out in the Cooperative Agreement with DOE for this Clean Coal Technology project. Overall plant availability (defined as the percentage of time that the LPMEOH{trademark} demonstration unit was able to operate, with the exclusion of scheduled outages) was 97.5%, and the longest operating run without interruption of any kind was 94 days. Over 103.9 million gallons of methanol was produced; Eastman accepted all of the available methanol for use in the production of methyl acetate, and ultimately cellulose acetate and acetic acid.

  8. DOWNSTREAM IMPACTS OF SLUDGE MASS REDUCTION VIA ALUMINUM DISSOLUTION ON DWPF PROCESSING OF SAVANNAH RIVER SITE HIGH LEVEL WASTE - 9382

    SciTech Connect (OSTI)

    Pareizs, J; Cj Bannochie, C; Michael Hay, M; Daniel McCabe, D

    2009-01-14T23:59:59.000Z

    The SRS sludge that was to become a major fraction of Sludge Batch 5 (SB5) for the Defense Waste Processing Facility (DWPF) contained a large fraction of H-Modified PUREX (HM) sludge, containing a large fraction of aluminum compounds that could adversely impact the processing and increase the vitrified waste volume. It is beneficial to reduce the non-radioactive fraction of the sludge to minimize the number of glass waste canisters that must be sent to a Federal Repository. Removal of aluminum compounds, such as boehmite and gibbsite, from sludge can be performed with the addition of NaOH solution and heating the sludge for several days. Preparation of SB5 involved adding sodium hydroxide directly to the waste tank and heating the contents to a moderate temperature through slurry pump operation to remove a fraction of this aluminum. The Savannah River National Laboratory (SRNL) was tasked with demonstrating this process on actual tank waste sludge in our Shielded Cells Facility. This paper evaluates some of the impacts of aluminum dissolution on sludge washing and DWPF processing by comparing sludge processing with and without aluminum dissolution. It was necessary to demonstrate these steps to ensure that the aluminum removal process would not adversely impact the chemical and physical properties of the sludge which could result in slower processing or process upsets in the DWPF.

  9. Process Knowledge Summary Report for Advanced Test Reactor Complex Contact-Handled Transuranic Waste Drum TRA010029

    SciTech Connect (OSTI)

    B. R. Adams; R. P. Grant; P. R. Smith; J. L. Weisgerber

    2013-09-01T23:59:59.000Z

    This Process Knowledge Summary Report summarizes information collected to satisfy the transportation and waste acceptance requirements for the transfer of one drum containing contact-handled transuranic (TRU) actinide standards generated by the Idaho National Laboratory at the Advanced Test Reactor (ATR) Complex to the Advanced Mixed Waste Treatment Project (AMWTP) for storage and subsequent shipment to the Waste Isolation Pilot Plant for final disposal. The drum (i.e., Integrated Waste Tracking System Bar Code Number TRA010029) is currently stored at the Materials and Fuels Complex. The information collected includes documentation that addresses the requirements for AMWTP and applicable sections of their Resource Conservation and Recovery Act permits for receipt and disposal of this TRU waste generated from ATR. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for this TRU waste originating from ATR.

  10. DWPF (Defense Waste Processing Facility) canister impact testing and analyses for the Transportation Technology Center

    SciTech Connect (OSTI)

    Farnsworth, R.K.; Mishima, J.

    1988-12-01T23:59:59.000Z

    A legal weight truck cask design has been developed for the US Department of Energy by GA Technologies, Inc. The cask will be used to transport defense high-level waste canisters produced by the Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The development of the cask required the collection of impact data for the DWPF canisters. The Materials Characterization Center (MCC) performed this work under the guidance of the Transportation Technology Center (TTC) at Sandia National Laboratories. Two full-scale DWPF canisters filled with nonradioactive borosilicate glass were impacted under ''normal'' and ''hypothetical'' accident conditions. Two canisters, supplied by the DWPF, were tested. Each canister was vertically dropped on the bottom end from a height of either 0.3 m or 9.1 m (for normal or hypothetical accident conditions, respectively). The structural integrity of each canister was then examined using helium leak and dye penetrant testing. The canisters' diameters and heights, which had been previously measured, were then remeasured to determine how the canister dimensions had changed. Following structural integrity testing, the canisters were flaw leak tested. For transportation flaw leak testing, four holes were fabricated into the shell of canister A-27 (0.3 m drop height). The canister was then transported a total distance of 2069 miles. During transport, the waste form material that fell from each flaw was collected to determine the amount of size distribution of each flaw release. 2 refs., 8 figs., 12 tabs.

  11. Process for the elimination of waste water produced upon the desulfurization of coking oven gas by means of wash solution containing organic oxygen-carrier, with simultaneous recovery of elemental sulfur

    SciTech Connect (OSTI)

    Diemer, P.; Brake, W.; Dittmer, R.

    1985-04-16T23:59:59.000Z

    A process is disclosed for the elimination of waste water falling out with the desulfurization of coking oven gas by means of an organic oxygen carrier-containing washing solution with simultaneous recovery of elemental sulfur. The waste water is decomposed in a combustion chamber in a reducing atmosphere at temperatures between about 1000/sup 0/ and 1100/sup 0/ C. under such conditions that the mole ratio of H/sub 2/S:SO/sub 2/ in the exhaust gas of the combustion chamber amounts to at least 2:1. Sulfur falling out is separated and the sensible heat of the exhaust gas is utilized for steam generation. The cooled and desulfurized exhaust gas is added to the coking oven gas before the pre-cooling. Sulfur falling out from the washing solution in the oxidizer is separated out and lead into the combustion chamber together with the part of the washing solution discharged as waste water from the washing solution circulation. Preferred embodiments include that the sulfur loading of the waste water can amount to up to about 370 kg sulfur per m/sup 3/ waste water; having the cooling of sulfur-containing exhaust gas leaving the combustion chamber follow in a waste heat boiler and a sulfur condenser heated by pre-heated boiler feed water, from which condenser sulfur is discharged in liquid state.

  12. MODELING OF PLUTONIUM RECOVERY AND DISCARD PROCESSES FOR THE PURPOSE OF SELECTING OPTIMUM (MINIMUM WASTE, COST AND DOSE) RESIDUE DISPOSITIONS

    SciTech Connect (OSTI)

    M. A. ROBINSON; M. B. KINKER; ET AL

    2001-04-01T23:59:59.000Z

    Researchers have developed a quantitative basis for disposition of actinide-bearing process residues. Research included the development of a technical rationale for determining when residues could be considered unattractive for proliferation purposes, and establishing plutonium-concentration-based discard ceilings of unimmobilized residues and richer discard ceilings for immobilized monolithic waste forms. Further quantitative analysis (process modeling) identifies the plutonium (Pu) concentration at which residues should be discarded to immobilization in order to minimize the quantifiable negative consequences of residue processing (cost, waste, dose). Results indicate that optimum disposition paths can be identified by process modeling, and that across-the-board discard decisions maximize negative consequences.

  13. Characterization of decontamination and decommissioning wastes expected from the major processing facilities in the 200 Areas

    SciTech Connect (OSTI)

    Amato, L.C.; Franklin, J.D.; Hyre, R.A.; Lowy, R.M.; Millar, J.S.; Pottmeyer, J.A. [Los Alamos Technical Associates, Kennewick, WA (United States); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (United States)

    1994-08-01T23:59:59.000Z

    This study was intended to characterize and estimate the amounts of equipment and other materials that are candidates for removal and subsequent processing in a solid waste facility when the major processing and handling facilities in the 200 Areas of the Hanford Site are decontaminated and decommissioned. The facilities in this study were selected based on processing history and on the magnitude of the estimated decommissioning cost cited in the Surplus Facilities Program Plan; Fiscal Year 1993 (Winship and Hughes 1992). The facilities chosen for this study include B Plant (221-B), T Plant (221-T), U Plant (221-U), the Uranium Trioxide (UO{sub 3}) Plant (224-U and 224-UA), the Reduction Oxidation (REDOX) or S Plant (202-S), the Plutonium Concentration Facility for B Plant (224-B), and the Concentration Facility for the Plutonium Finishing Plant (PFP) and REDOX (233-S). This information is required to support planning activities for current and future solid waste treatment, storage, and disposal operations and facilities.

  14. IMPACTS OF ANTIFOAM ADDITIONS AND ARGON BUBBLING ON DEFENSE WASTE PROCESSING FACILITY REDUCTION/OXIDATION

    SciTech Connect (OSTI)

    Jantzen, C.; Johnson, F.

    2012-06-05T23:59:59.000Z

    During melting of HLW glass, the REDOX of the melt pool cannot be measured. Therefore, the Fe{sup +2}/{Sigma}Fe ratio in the glass poured from the melter must be related to melter feed organic and oxidant concentrations to ensure production of a high quality glass without impacting production rate (e.g., foaming) or melter life (e.g., metal formation and accumulation). A production facility such as the Defense Waste Processing Facility (DWPF) cannot wait until the melt or waste glass has been made to assess its acceptability, since by then no further changes to the glass composition and acceptability are possible. therefore, the acceptability decision is made on the upstream process, rather than on the downstream melt or glass product. That is, it is based on 'feed foward' statistical process control (SPC) rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. Use of the DWPF REDOX model has controlled the balanjce of feed reductants and oxidants in the Sludge Receipt and Adjustment Tank (SRAT). Once the alkali/alkaline earth salts (both reduced and oxidized) are formed during reflux in the SRAT, the REDOX can only change if (1) additional reductants or oxidants are added to the SRAT, the Slurry Mix Evaporator (SME), or the Melter Feed Tank (MFT) or (2) if the melt pool is bubble dwith an oxidizing gas or sparging gas that imposes a different REDOX target than the chemical balance set during reflux in the SRAT.

  15. Fluid bed gasification – Plasma converter process generating energy from solid waste: Experimental assessment of sulphur species

    SciTech Connect (OSTI)

    Morrin, Shane, E-mail: shane.morrin@ucl.ac.uk [Department of Chemical Engineering, University College London, London WC1E 7JE (United Kingdom); Advanced Plasma Power, Swindon, Wiltshire SN3 4DE (United Kingdom); Lettieri, Paola, E-mail: p.lettieri@ucl.ac.uk [Department of Chemical Engineering, University College London, London WC1E 7JE (United Kingdom); Chapman, Chris, E-mail: chris.chapman@app-uk.com [Advanced Plasma Power, Swindon, Wiltshire SN3 4DE (United Kingdom); Taylor, Richard, E-mail: richard.taylor@app-uk.com [Advanced Plasma Power, Swindon, Wiltshire SN3 4DE (United Kingdom)

    2014-01-15T23:59:59.000Z

    Highlights: • We investigate gaseous sulphur species whilst gasifying sulphur-enriched wood pellets. • Experiments performed using a two stage fluid bed gasifier – plasma converter process. • Notable SO{sub 2} and relatively low COS levels were identified. • Oxygen-rich regions of the bed are believed to facilitate SO{sub 2}, with a delayed release. • Gas phase reducing regions above the bed would facilitate more prompt COS generation. - Abstract: Often perceived as a Cinderella material, there is growing appreciation for solid waste as a renewable content thermal process feed. Nonetheless, research on solid waste gasification and sulphur mechanisms in particular is lacking. This paper presents results from two related experiments on a novel two stage gasification process, at demonstration scale, using a sulphur-enriched wood pellet feed. Notable SO{sub 2} and relatively low COS levels (before gas cleaning) were interesting features of the trials, and not normally expected under reducing gasification conditions. Analysis suggests that localised oxygen rich regions within the fluid bed played a role in SO{sub 2}’s generation. The response of COS to sulphur in the feed was quite prompt, whereas SO{sub 2} was more delayed. It is proposed that the bed material sequestered sulphur from the feed, later aiding SO{sub 2} generation. The more reducing gas phase regions above the bed would have facilitated COS – hence its faster response. These results provide a useful insight, with further analysis on a suite of performed experiments underway, along with thermodynamic modelling.

  16. Selection of Steady-State Process Simulation Software to Optimize Treatment of Radioactive and Hazardous Waste

    SciTech Connect (OSTI)

    Nichols, T. T.; Barnes, C. M.; Lauerhass, L.; Taylor, D. D.

    2001-06-01T23:59:59.000Z

    The process used for selecting a steady-state process simulator under conditions of high uncertainty and limited time is described. Multiple waste forms, treatment ambiguity, and the uniqueness of both the waste chemistries and alternative treatment technologies result in a large set of potential technical requirements that no commercial simulator can totally satisfy. The aim of the selection process was two-fold. First, determine the steady-state simulation software that best, albeit not completely, satisfies the requirements envelope. And second, determine if the best is good enough to justify the cost. Twelve simulators were investigated with varying degrees of scrutiny. The candidate list was narrowed to three final contenders: ASPEN Plus 10.2, PRO/II 5.11, and CHEMCAD 5.1.0. It was concluded from ''road tests'' that ASPEN Plus appears to satisfy the project's technical requirements the best and is worth acquiring. The final software decisions provide flexibility: they involve annual rather than multi-year licensing, and they include periodic re-assessment.

  17. Feed Composition for Sodium-Bearing Waste Treatment Process, Rev. 3

    SciTech Connect (OSTI)

    Barnes, Charles Marshall

    2003-09-01T23:59:59.000Z

    Treatment of sodium-bearing waste (SBW) at the Idaho Nuclear Technology and Engineering Center (INTEC) within the Idaho National Engineering and Environmental Laboratory is mandated by a Settlement Agreement between the Department of Energy and the State of Idaho. One of the requirements of the Settlement Agreement is to complete treatment of SBW by December 31, 2012. To support both design and development studies for the SBW treatment process, detailed feed compositions are needed. This report contains the expected compositions of these feed streams and the sources and methods used in obtaining these compositions.

  18. Fabrication of hollow mesoporous NiO hexagonal microspheres via hydrothermal process in ionic liquid

    SciTech Connect (OSTI)

    Zhao, Jinbo, E-mail: zhaojinb@gmail.com [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (Ministry of Education), Shandong University, 250061, Jinan (China) [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (Ministry of Education), Shandong University, 250061, Jinan (China); School of Materials Science and Engineering, Shandong University, 250061, Jinan (China); Wu, Lili, E-mail: wulili@sdu.edu.cn [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (Ministry of Education), Shandong University, 250061, Jinan (China) [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (Ministry of Education), Shandong University, 250061, Jinan (China); School of Materials Science and Engineering, Shandong University, 250061, Jinan (China); Zou, Ke, E-mail: zouk2005@163.com [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (Ministry of Education), Shandong University, 250061, Jinan (China) [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (Ministry of Education), Shandong University, 250061, Jinan (China); School of Materials Science and Engineering, Shandong University, 250061, Jinan (China)

    2011-12-15T23:59:59.000Z

    Highlights: Black-Right-Pointing-Pointer Ni(OH){sub 2} precursors were synthesized in ionic liquid and water solution by hydrothermal method. Black-Right-Pointing-Pointer NiO hollow microspheres were prepared by thermal treatment of Ni(OH){sub 2} precursors. Black-Right-Pointing-Pointer NiO hollow microspheres were self-assembled by mesoporous cubic and hexagonal nanocrystals with high specific surface area. Black-Right-Pointing-Pointer The mesoporous structure is stable at 773 K. Black-Right-Pointing-Pointer The ionic liquid absorbed on the O-terminate surface of the crystals to form hydrogen bond and played key roles in determining the final shape of the NiO novel microstructure. -- Abstract: The novel NiO hexagonal hollow microspheres have been successfully prepared by annealing Ni(OH){sub 2}, which was synthesized via an ionic liquid-assisted hydrothermal method. The samples were characterized by X-ray diffraction (XRD), field emission scanning electron microscope (FE-SEM), transmission electron microscopy (TEM), N{sub 2} adsorption-desorption and Fourier transform infrared spectrometer (FTIR). The results show that the hollow NiO microstructures are self-organized by mesoporous cubic and hexagonal nanocrystals. The mesoporous structure possessed good thermal stability and high specific surface area (ca. 83 m{sup 2}/g). The ionic liquid 1-butyl-3methylimidazolium tetrafluoroborate ([Bmim][BF{sub 4}]) was found to play a key role in controlling the morphology of NiO microstructures during the hydrothermal process. The special hollow mesoporous architectures will have potential applications in many fields, such as catalysts, absorbents, sensors, drug-delivery carriers, acoustic insulators and supercapacitors.

  19. HIGH LEVEL WASTE (HLW) VITRIFICATION EXPERIENCE IN THE US: APPLICATION OF GLASS PRODUCT/PROCESS CONTROL TO OTHERHLW AND HAZARDOUS WASTES

    SciTech Connect (OSTI)

    Jantzen, C; James Marra, J

    2007-09-17T23:59:59.000Z

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. At the Savannah River Site (SRS) actual HLW tank waste has successfully been processed to stringent product and process constraints without any rework into a stable borosilicate glass waste since 1996. A unique 'feed forward' statistical process control (SPC) has been used rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. In SQC, the glass product is sampled after it is vitrified. Individual glass property models form the basis for the 'feed forward' SPC. The property models transform constraints on the melt and glass properties into constraints on the feed composition. The property models are mechanistic and depend on glass bonding/structure, thermodynamics, quasicrystalline melt species, and/or electron transfers. The mechanistic models have been validated over composition regions well outside of the regions for which they were developed because they are mechanistic. Mechanistic models allow accurate extension to radioactive and hazardous waste melts well outside the composition boundaries for which they were developed.

  20. Commercial-Scale Demonstration of the Liquid Phase methanol (LPMEOH) Process A DOE Assessment

    SciTech Connect (OSTI)

    National Energy Technology Laboratory

    2003-10-27T23:59:59.000Z

    The U.S. Department of Energy (DOE) Clean Coal Technology (CCT) Program seeks to offer the energy marketplace more efficient and environmentally benign coal utilization technology options by demonstrating them in industrial settings. This document is a DOE post-project assessment (PPA) of one of the projects selected in Round III of the CCT Program, the commercial-scale demonstration of the Liquid Phase Methanol (LPMEOH{trademark}) Process, initially described in a Report to Congress by DOE in 1992. Methanol is an important, large-volume chemical with many uses. The desire to demonstrate a new process for the production of methanol from coal, prompted Air Products and Chemicals, Inc. (Air Products) to submit a proposal to DOE. In October 1992, DOE awarded a cooperative agreement to Air Products to conduct this project. In March 1995, this cooperative agreement was transferred to Air Products Liquid Phase Conversion Company, L.P. (the Partnership), a partnership between Air Products and Eastman Chemical Company (Eastman). DOE provided 43 percent of the total project funding of $213.7 million. Operation of the LPMEOH Demonstration Unit, which is sited at Eastman's chemicals-from-coal complex in Kingsport, Tennessee, commenced in April 1997. Although operation of the CCT project was completed in December 2002, Eastman continues to operate the LPMEOH Demonstration Unit for the production of methanol. The independent evaluation contained herein is based primarily on information from Volume 2 of the project's Final Report (Air Products Liquid Phase Conversion Co., L.P. 2003), as well as other references cited.

  1. Structural Integrity Program for the 300,000-Gallon Radioactive Liquid Waste Storage Tanks at the Idaho Nuclear Technology and Engineering Center

    SciTech Connect (OSTI)

    Bryant, Jeffrey W.

    2010-08-12T23:59:59.000Z

    This report provides a record of the Structural Integrity Program for the 300,000-gal liquid waste storage tanks and associated equipment at the Idaho Nuclear Technology and Engineering Center, as required by U.S. Department of Energy M 435.1-1, “Radioactive Waste Management Manual.” This equipment is known collectively as the Tank Farm Facility. This report is an update, and replaces the previous report by the same title issued April 2003. The conclusion of this report is that the Tank Farm Facility tanks, vaults, and transfer systems that remain in service for storage are structurally adequate, and are expected to remain structurally adequate over the remainder of their planned service life through 2012. Recommendations are provided for continued monitoring of the Tank Farm Facility.

  2. The nucleation process and the roles of structure and density fluctuations in supercooled liquid Fe

    SciTech Connect (OSTI)

    Li, Rong; Wu, Yongquan, E-mail: yqwu@shu.edu.cn; Xiao, Junjiang [Shanghai Key Laboratory of Modern Metallurgy and Materials Processing, Shanghai University, Shanghai 200072 (China)] [Shanghai Key Laboratory of Modern Metallurgy and Materials Processing, Shanghai University, Shanghai 200072 (China)

    2014-01-21T23:59:59.000Z

    We observed homogeneous nucleation process of supercooled liquid Fe by molecular dynamics simulations. Using bond-orientational order parameters together with Voronoi polyhedron method, we characterized local structure, calculated the volume of Voronoi polyhedra of atoms and identified the structure and density fluctuations. We monitored the formation of nucleus and analyzed its inner structure. The birth and growth of the pre-nucleus and nucleus are accompanied with aggregating and disaggregating processes in the time scale of femtosecond. Only the initial solid-like clusters (ISLC), ranging from 1 to 7 atoms, pop up directly from liquid. The relation between the logarithm of number of clusters and the cluster size was found to be linear for ISLCs and was observed to be parabolic for all solid-like clusters (SLC) due to aggregating and disaggregating effects. The nucleus and pre-nuclei mainly consist of body centered cubic (BCC) and hexagonal close packed atoms, while the BCC atoms tend to be located at the surface. Medium-range structure fluctuations induce the birth of ISLCs, benefit the aggregation of embryos and remarkably promote the nucleation. But density fluctuations contribute little to nucleation. The lifetime of most icosahedral-like atoms (ICO) is shorter than 0.7 ps. No obvious relationship was found between structure/density fluctuations and the appearance of ICO atoms.

  3. The Mixed Waste Management Facility: Technology selection and implementation plan, Part 2, Support processes

    SciTech Connect (OSTI)

    Streit, R.D.; Couture, S.A.

    1995-03-01T23:59:59.000Z

    The purpose of this document is to establish the foundation for the selection and implementation of technologies to be demonstrated in the Mixed Waste Management Facility, and to select the technologies for initial pilot-scale demonstration. Criteria are defined for judging demonstration technologies, and the framework for future technology selection is established. On the basis of these criteria, an initial suite of technologies was chosen, and the demonstration implementation scheme was developed. Part 1, previously released, addresses the selection of the primary processes. Part II addresses process support systems that are considered ``demonstration technologies.`` Other support technologies, e.g., facility off-gas, receiving and shipping, and water treatment, while part of the integrated demonstration, use best available commercial equipment and are not selected against the demonstration technology criteria.

  4. Evaluation of gasification and novel thermal processes for the treatment of municipal solid waste

    SciTech Connect (OSTI)

    Niessen, W.R.; Marks, C.H.; Sommerlad, R.E. [Camp Dresser and McKee, Inc., Cambridge, MA (United States)] [Camp Dresser and McKee, Inc., Cambridge, MA (United States)

    1996-08-01T23:59:59.000Z

    This report identifies seven developers whose gasification technologies can be used to treat the organic constituents of municipal solid waste: Energy Products of Idaho; TPS Termiska Processor AB; Proler International Corporation; Thermoselect Inc.; Battelle; Pedco Incorporated; and ThermoChem, Incorporated. Their processes recover heat directly, produce a fuel product, or produce a feedstock for chemical processes. The technologies are on the brink of commercial availability. This report evaluates, for each technology, several kinds of issues. Technical considerations were material balance, energy balance, plant thermal efficiency, and effect of feedstock contaminants. Environmental considerations were the regulatory context, and such things as composition, mass rate, and treatability of pollutants. Business issues were related to likelihood of commercialization. Finally, cost and economic issues such as capital and operating costs, and the refuse-derived fuel preparation and energy conversion costs, were considered. The final section of the report reviews and summarizes the information gathered during the study.

  5. Study on LiCl waste salt treatment process by layer melt crystallization

    SciTech Connect (OSTI)

    Cho, Yung-Zun; Lee, Tae-Kyo; Choi, Jung-Hoon; Eun, Hee-Chul; Park, Hwan-Seo; Kim, In-Tae; Park, Geun-Il [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    Layer melt crystallization operated in a static mode has been applied to separate Group I and II chlorides from surrogate LiCl waste salt. The effects of operating conditions such as crystal growing rate(or flux) and initial impurity concentration on separation (or concentration) of cesium, strontium and barium involved in a LiCl melts were analyzed. In a layer crystallization process, separation was impaired by occlusion of impurities and by residual melt adhering to LiCl crystal after at the end of the process. The crystal growth rate strongly affects the crystal structure, therefore the separation efficiency, while the effect of the initial Cs and Sr concentration in LiCl molten salt was nearly negligible. (authors)

  6. Evaluation of alternative chemical additives for high-level waste vitrification feed preparation processing

    SciTech Connect (OSTI)

    Seymour, R.G.

    1995-06-07T23:59:59.000Z

    During the development of the feed processing flowsheet for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), research had shown that use of formic acid (HCOOH) could accomplish several processing objectives with one chemical addition. These objectives included the decomposition of tetraphenylborate, chemical reduction of mercury, production of acceptable rheological properties in the feed slurry, and controlling the oxidation state of the glass melt pool. However, the DEPF research had not shown that some vitrification slurry feeds had a tendency to evolve hydrogen (H{sub 2}) and ammonia (NH{sub 3}) as the result of catalytic decomposition of CHOOH with noble metals (rhodium, ruthenium, palladium) in the feed. Testing conducted at Pacific Northwest Laboratory and later at the Savannah River Technical Center showed that the H{sub 2} and NH{sub 3} could evolve at appreciable rates and quantities. The explosive nature of H{sub 2} and NH{sub 3} (as ammonium nitrate) warranted significant mitigation control and redesign of both facilities. At the time the explosive gas evolution was discovered, the DWPF was already under construction and an immediate hardware fix in tandem with flowsheet changes was necessary. However, the Hanford Waste Vitrification Plant (HWVP) was in the design phase and could afford to take time to investigate flowsheet manipulations that could solve the problem, rather than a hardware fix. Thus, the HWVP began to investigate alternatives to using HCOOH in the vitrification process. This document describes the selection, evaluation criteria, and strategy used to evaluate the performance of the alternative chemical additives to CHOOH. The status of the evaluation is also discussed.

  7. In Situ Grouting of Liquid Waste Disposal Trenches and Experimental Reactor Fuel Disposal Wells at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Johnson, Ch.; Cange, J.; Lambert, R. [Bechtel Jacobs Company, LLC, Oak Ridge, TN (United States); Trujillo, E. [BWXT Pantex, LLC, Amarillo, TX (United States); Julius, J. [U.S. DOE, Oak Ridge Operations Office, Oak Ridge, TN (United States)

    2008-07-01T23:59:59.000Z

    In the early to mid-1960's, liquid low-level wastes (LLLW) generated at Oak Ridge National Laboratory were disposed of in specially-constructed, gravel-filled trenches within the Melton Valley watershed at the lab. The initial selected remedy for Trenches 5 and 7 was in situ vitrification; however, an amendment to the record of decision changed the remedy to in situ grouting of the trenches. The work was accomplished by filling the void space within the crushed stone section of each trench with cementitious grout. The contaminated soil surrounding the trenches (1-m perimeter) was then grouted with acrylamide grout. At the HRE fuel wells, a 1-m ring of soil surrounding the fuel wells was grouted with acrylamide. The results of the hydraulic conductivity tests ranged from 4.74 x 10{sup -6} to 3.60 x 10{sup -7} cm/sec, values that were well below the 1 x 10{sup -5} cm/sec design criterion. In summary: The ISG Project was conducted to decrease hydraulic conductivity and thereby decrease water flow and contaminate migration from the area of the trenches. The initial remedy for Trenches 5 and 7 in the Melton Valley ROD was for in situ vitrification of the trench matrix. The remedy was changed to in situ grouting of the trenches and HRE fuel wells through an amendment to the ROD after moisture was found in the trenches. The grouting of the trenches was accomplished by filling the void space within the crushed stone section of each trench with cementitious grout. The contaminated soil surrounding the trenches (1-m perimeter) was then grouted with acrylamide grout to further reduce water infiltration. Soil backfill above each of the seven HRE fuel wells was removed to a depth of approximately 1 m by augering, and the soils were replaced with a cement plug to prevent water infiltration from migrating down the original borehole. Soil surrounding the fuel wells was then grouted with acrylamide to ensure water infiltration through the HRE fuel wells is prevented. A summary of the quantities used is shown. After completion of grouting, in-situ hydraulic conductivities of the grouted materials were measured to verify attainment of the design objective. The areas were then covered with multi-layer caps as part of the MV hydrologic isolation project. (authors)

  8. TECHNICAL ASSESSMENT OF BULK VITRIFICATION PROCESS & PRODUCT FOR TANK WASTE TREATMENT AT THE DEPARTMENT OF ENERGY HANFORD SITE

    SciTech Connect (OSTI)

    SCHAUS, P.S.

    2006-07-21T23:59:59.000Z

    At the U.S. Department of Energy (DOE) Hanford Site, the Waste Treatment Plant (WTP) is being constructed to immobilize both high-level waste (IUW) for disposal in a national repository and low-activity waste (LAW) for onsite, near-surface disposal. The schedule-controlling step for the WTP Project is vitrification of the large volume of LAW, current capacity of the WTP (as planned) would require 50 years to treat the Hanford tank waste, if the entire LAW volume were to be processed through the WTP. To reduce the time and cost for treatment of Hanford Tank Waste, and as required by the Tank Waste Remediation System Environmental Impact Statement Record of Decision and the Hanford Federal Facility Consent Agreement (Tn-Party Agreement), DOE plans to supplement the LAW treatment capacity of the WTP. Since 2002, DOE, in cooperation with the Environmental Protection Agency and State of Washington Department of Ecology has been evaluating technologies that could provide safe and effective supplemental treatment of LAW. Current efforts at Hanford are intended to provide additional information to aid a joint agency decision on which technology will be used to supplement the WTP. A Research, Development and Demonstration permit has been issued by the State of Washington to build and (for a limited time) operate a Demonstration Bulk Vitrification System (DBVS) facility to provide information for the decision on a supplemental treatment technology for up to 50% of the LAW. In the Bulk Vitrification (BV) process, LAW, soil, and glass-forming chemicals are mixed, dried, and placed in a refractory-lined box, Electric current, supplied through two graphite electrodes in the box, melts the waste feed, producing a durable glass waste-form. Although recent modifications to the process have resulted in significant improvements, there are continuing technical concerns.

  9. VERIFICATION OF THE DEFENSE WASTE PROCESSING FACILITY'S (DWPF) PROCESS DIGESTION METHOD FOR THE SLUDGE BATCH 7A QUALIFICATION SAMPLE

    SciTech Connect (OSTI)

    Click, D.; Edwards, T.; Jones, M.; Wiedenman, B.

    2011-03-14T23:59:59.000Z

    For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs confirmation of the applicability of the digestion method to be used by the DWPF lab for elemental analysis of Sludge Receipt and Adjustment Tank (SRAT) receipt samples and SRAT product process control samples. DWPF SRAT samples are typically dissolved using a room temperature HF-HNO{sub 3} acid dissolution (i.e., DWPF Cold Chem Method, see DWPF Procedure SW4-15.201) and then analyzed by inductively coupled plasma - atomic emission spectroscopy (ICP-AES). This report contains the results and comparison of data generated from performing the Aqua Regia (AR), Sodium peroxide/Hydroxide Fusion (PF) and DWPF Cold Chem (CC) method digestions of Sludge Batch 7a (SB7a) SRAT Receipt and SB7a SRAT Product samples. The SB7a SRAT Receipt and SB7a SRAT Product samples were prepared in the SRNL Shielded Cells, and the SRAT Receipt material is representative of the sludge that constituates the SB7a Batch or qualification composition. This is the sludge in Tank 51 that is to be transferred into Tank 40, which will contain the heel of Sludge Batch 6 (SB6), to form the Sb7a Blend composition.

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

    SciTech Connect (OSTI)

    Amoroso, J. [Savannah River National Laboratory, Aiken, SC (United States); Marra, J. [Savannah River National Laboratory, Aiken, SC (United States)

    2014-10-02T23:59:59.000Z

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

  11. State waste discharge permit application: 200 Area Treated Effluent Disposal Facility (Project W-049H)

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    As part of the original Hanford Federal Facility Agreement and Concent Order negotiations, US DOE, US EPA and the Washington State Department of Ecology agreed that liquid effluent discharges to the ground to the Hanford Site are subject to permitting in the State Waste Discharge Permit Program (SWDP). This document constitutes the SWDP Application for the 200 Area TEDF stream which includes the following streams discharged into the area: Plutonium Finishing Plant waste water; 222-S laboratory Complex waste water; T Plant waste water; 284-W Power Plant waste water; PUREX chemical Sewer; B Plant chemical sewer, process condensate, steam condensate; 242-A-81 Water Services waste water.

  12. The production of fuels and chemicals from food processing wastes using a novel fermenter separator

    SciTech Connect (OSTI)

    Dale, M.C.; Venkatesh, K.V.; Choi, Hojoon; Moelhman, M.; Saliceti, L.; Okos, M.R.; Wankat, P.C.

    1991-12-01T23:59:59.000Z

    During 1991, considerable progress was made on the waste utilization project. Two small Wisconsin companies have expressed an interest in promoting and developing the ICRS technology. Pilot plant sites at (1) Hopkinton, IA, for a sweet whey plant, and Beaver Dam WI, for an acid whey site have been under development siting ICRS operations. The Hopkinton, IA site is owned and operated by Permeate Refining Inc., who have built a batch ethanol plant across the street from Swiss Valley Farms cheddar cheese operations. Permeate from Swiss Valley is piped across to PRI. PRI has signed a contract to site a 300--500,000 gallon/yr to ICRS pilot plant. They feel that the lower labor, lower energy, continuous process offered by the ICRS will substantially improve their profitability. Catalytics, Inc, is involved with converting whey from a Kraft cream cheese operation to ethanol and yeast. A complete project including whey concentration, sterilization, and yeast growth has been designed for this site. Process design improvements with the ICRS focussed on ethanol recovery techniques during this year's project. A solvent absorption/extractive distillation (SAED) process has been developed which offers the capability of obtaining an anhydrous ethanol product from vapors off 3 to 9% ethanol solutions using very little energy for distillation. Work on products from waste streams was also performed. a. Diacetyl as a high value flavor compound was very successfully produced in a Stirred Tank Reactor w/Separation. b. Yeast production from secondary carbohydrates in the whey, lactic acid, and glycerol was studied. c. Lactic acid production from cellulose and lactose studies continued. d. Production of anti-fungal reagents by immobilized plant cells; Gossypol has antifungal properties and is produced by G. arboretum.

  13. SUMMARY OF FY11 SULFATE RETENTION STUDIES FOR DEFENSE WASTE PROCESSING FACILITY GLASS

    SciTech Connect (OSTI)

    Fox, K.; Edwards, T.

    2012-05-08T23:59:59.000Z

    This report describes the results of studies related to the incorporation of sulfate in high level waste (HLW) borosilicate glass produced at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). A group of simulated HLW glasses produced for earlier sulfate retention studies was selected for full chemical composition measurements to determine whether there is any clear link between composition and sulfate retention over the compositional region evaluated. In addition, the viscosity of several glasses was measured to support future efforts in modeling sulfate solubility as a function of predicted viscosity. The intent of these studies was to develop a better understanding of sulfate retention in borosilicate HLW glass to allow for higher loadings of sulfate containing waste. Based on the results of these and other studies, the ability to improve sulfate solubility in DWPF borosilicate glasses lies in reducing the connectivity of the glass network structure. This can be achieved, as an example, by increasing the concentration of alkali species in the glass. However, this must be balanced with other effects of reduced network connectivity, such as reduced viscosity, potentially lower chemical durability, and in the case of higher sodium and aluminum concentrations, the propensity for nepheline crystallization. Future DWPF processing is likely to target higher waste loadings and higher sludge sodium concentrations, meaning that alkali concentrations in the glass will already be relatively high. It is therefore unlikely that there will be the ability to target significantly higher total alkali concentrations in the glass solely to support increased sulfate solubility without the increased alkali concentration causing failure of other Product Composition Control System (PCCS) constraints, such as low viscosity and durability. No individual components were found to provide a significant improvement in sulfate retention (i.e., an increase of the magnitude necessary to have a dramatic impact on blending, washing, or waste loading strategies for DWPF) for the glasses studied here. In general, the concentrations of those species that significantly improve sulfate solubility in a borosilicate glass must be added in relatively large concentrations (e.g., 13 to 38 wt % or more of the frit) in order to have a substantial impact. For DWPF, these concentrations would constitute too large of a portion of the frit to be practical. Therefore, it is unlikely that specific additives may be introduced into the DWPF glass via the frit to significantly improve sulfate solubility. The results presented here continue to show that sulfate solubility or retention is a function of individual glass compositions, rather than a property of a broad glass composition region. It would therefore be inappropriate to set a single sulfate concentration limit for a range of DWPF glass compositions. Sulfate concentration limits should continue to be identified and implemented for each sludge batch. The current PCCS limit is 0.4 wt % SO{sub 4}{sup 2-} in glass, although frit development efforts have led to an increased limit of 0.6 wt % for recent sludge batches. Slightly higher limits (perhaps 0.7-0.8 wt %) may be possible for future sludge batches. An opportunity for allowing a higher sulfate concentration limit at DWPF may lay lie in improving the laboratory experiments used to set this limit. That is, there are several differences between the crucible-scale testing currently used to define a limit for DWPF operation and the actual conditions within the DWPF melter. In particular, no allowance is currently made for sulfur partitioning (volatility versus retention) during melter processing as the sulfate limit is set for a specific sludge batch. A better understanding of the partitioning of sulfur in a bubbled melter operating with a cold cap as well as the impacts of sulfur on the off-gas system may allow a higher sulfate concentration limit to be established for the melter feed. This approach would have to be taken carefully to ensure that a

  14. Lessons learned from the EG&G consolidated hazardous waste subcontract and ESH&Q liability assessment process

    SciTech Connect (OSTI)

    Fix, N.J.

    1995-03-01T23:59:59.000Z

    Hazardous waste transportation, treatment, recycling, and disposal contracts were first consolidated at the Idaho National Engineering Laboratory in 1992 by EG&G Idaho, Inc. At that time, disposition of Resource, Conservation and Recovery Act hazardous waste, Toxic Substance Control Act waste, Comprehensive Environmental Response, Compensation, and Liability Act hazardous substances and contaminated media, and recyclable hazardous materials was consolidated under five subcontracts. The wastes were generated by five different INEL M&O contractors, under the direction of three different Department of Energy field offices. The consolidated contract reduced the number of facilities handling INEL waste from 27 to 8 qualified treatment, storage, and disposal facilities, with brokers specifically prohibited. This reduced associated transportation costs, amount and cost of contractual paperwork, and environmental liability exposure. EG&G reviewed this approach and proposed a consolidated hazardous waste subcontract be formed for the major EG&G managed DOE sites: INEL, Mound, Rocky Flats, Nevada Test Site, and 10 satellite facilities. After obtaining concurrence from DOE Headquarters, this effort began in March 1992 and was completed with the award of two master task subcontracts in October and November 1993. In addition, the effort included a team to evaluate the apparent awardee`s facilities for environment, safety, health, and quality (ESH&Q) and financial liability status. This report documents the evaluation of the process used to prepare, bid, and award the EG&G consolidated hazardous waste transportation, treatment, recycling, and/or disposal subcontracts and associated ESH&Q and financial liability assessments; document the strengths and weaknesses of the process; and propose improvements that would expedite and enhance the process for other DOE installations that used the process and for the re-bid of the consolidated subcontract, scheduled for 1997.

  15. Quasicrystalline Approach to Prediting the Spinel-Nepheline Liquidus: Application to Nuclear Waste Glass Processing

    SciTech Connect (OSTI)

    Jantzen, Carol

    2005-10-10T23:59:59.000Z

    The crystal-melt equilibria in complex fifteen component melts are modeled based on quasicrystalline concepts. A pseudobinary phase diagram between acmite (which melts incongruently to a transition metal ferrite spinel) and nepheline is defined. The pseudobinary lies within the Al{sub 2}O{sub 3}-Fe{sub 2}O{sub 3}-Na{sub 2}O-SiO{sub 2} quaternary system that defines the crystallization of basalt glass melts. The pseudobinary provides the partitioning of species between the melt and the primary liquidus phases. The medium range order of the melt and the melt-crystal exchange equilibria are defined based on a constrained mathematical treatment that considers the crystallochemical coordination of the elemental species in acmite and nepheline. The liquidus phases that form are shown to be governed by the melt polymerization and the octahedral site preference energies. This quasicrystalline liquidus model has been used to prevent unwanted crystallization in the world's largest high level waste (HLW) melter for the past three years while allowing >10 wt% higher waste loadings to be processed.

  16. Process for removing thorium and recovering vanadium from titanium chlorinator waste

    DOE Patents [OSTI]

    Olsen, Richard S. (Albany, OR); Banks, John T. (Corvallis, OR)

    1996-01-01T23:59:59.000Z

    A process for removal of thorium from titanium chlorinator waste comprising: (a) leaching an anhydrous titanium chlorinator waste in water or dilute hydrochloric acid solution and filtering to separate insoluble minerals and coke fractions from soluble metal chlorides; (b) beneficiating the insoluble fractions from step (a) on shaking tables to recover recyclable or otherwise useful TiO.sub.2 minerals and coke; and (c) treating filtrate from step (a) with reagents to precipitate and remove thorium by filtration along with acid metals of Ti, Zr, Nb, and Ta by the addition of the filtrate (a), a base and a precipitant to a boiling slurry of reaction products (d); treating filtrate from step (c) with reagents to precipitate and recover an iron vanadate product by the addition of the filtrate (c), a base and an oxidizing agent to a boiling slurry of reaction products; and (e) treating filtrate from step (d) to remove any remaining cations except Na by addition of Na.sub.2 CO.sub.3 and boiling.

  17. Technical assessment of processes to enable recycling of low-level contaminated metal waste

    SciTech Connect (OSTI)

    Reimann, G.A.

    1991-10-01T23:59:59.000Z

    Accumulations of metal waste exhibiting low levels of radioactivity (LLCMW) have become a national burden, both financially and environmentally. Much of this metal could be considered as a resource. The Department of Energy was assigned the task of inventorying and classifying LLCMW, identifying potential applications, and applying and/or developing the technology necessary to enable recycling. One application for recycled LLCMW is high-quality canisters for permanent repository storage of high-level waste (HLW). As many as 80,000 canisters will be needed by 2035. Much of the technology needed to decontaminate LLCMW has already been developed, but no integrated process has been described, even on a pilot scale, for recycling LLCMW into HLW canisters. This report reviews practices for removal of radionuclides and for producing low carbon stainless steel. Contaminants that readily form oxides may be reduced to below de minimis levels and combined with a slag. Most of the radioactivity remaining in the ingot is concentrated in the inclusions. Radionuclides that chemically resemble the elements that comprise stainless steel can not be removed effectively. Slag compositions, current melting practices, and canister fabrication techniques were reviewed.

  18. Enterprise Assessments Operational Awareness Record, Waste Treatment...

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

    system (LMH), the melter equipment support handling system (LSH), the radioactive solid waste handling system (RWH), and the radioactive liquid waste disposal system (RLD)....

  19. PILOT-SCALE TEST RESULTS OF A THIN FILM EVAPORATOR SYSTEM FOR MANAGEMENT OF LIQUID HIGH-LEVEL WASTES AT THE HANFORD SITE WASHINGTON USA -11364

    SciTech Connect (OSTI)

    CORBETT JE; TEDESCH AR; WILSON RA; BECK TH; LARKIN J

    2011-02-14T23:59:59.000Z

    A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORPIDOE), through Columbia Energy and Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper summarizes results of a pilot-scale test program conducted during calendar year 2010 as part of the ongoing technology maturation development scope for the WFE.

  20. Theoretical and Experimental Modelling of Bubble Formation with Connected Capillaries in Liquid Composite Moulding Processes

    E-Print Network [OSTI]

    Wielhorski, Yanneck; Bizet, Laurent; Bréard, Joël

    2012-01-01T23:59:59.000Z

    The void prediction in LCM processes sparks off interest within the composite material industry because it is a significant issue to keep the expected mechanical properties. The liquid properties, the preform geometry and the flow conditions impact the quantity of void entrapped inside the final product. The complex geometry of the reinforcement due to the arrangement of the bundles and the fibres is a key point to understand and quantify this phenomenon. This paper deals with both simple model networks which can occur inside a fabric representing connected capillaries, so-called "Pore Doublet Model (PDM)". A first is considering two capillaries converging on a node (T-junction) and a second is representing two capillaries interconnected with a supplying principle. These configurations can affect locally the evolution of flow fronts. First, experiments of bubble formed in a T-junction device have been performed and studied. Then a theoretical approach was proposed to forecast microvoid and macrovoid formation...

  1. Method for improving dissolution efficiency in gas-absorption and liquid extraction processes. [Patent application

    DOE Patents [OSTI]

    Kanak, B.E.; Stephenson, M.J.

    1980-01-11T23:59:59.000Z

    A method is described for improving dissolution efficiency in processes in which a feed fluid is introduced to a zone where it is contacted with a liquid solvent for preferentially removing a component of the feed and where part of the solvent so contacted undergoes transfer into the feed fluid to saturate the same. It has been found that such transfer significantly impairs dissolution efficiency. In accordance with the invention, an amount of the above-mentioned solvent is added to the feed fluid being introduced to the contact zone, the solvent being added in an amount sufficient to effect reduction or elimination of the above-mentioned transfer. Preferably, the solvent is added to the feed fluid in an amount saturating or supersaturating the feed fluid under the conditions prevailing in the contact zone.

  2. Process and apparatus for obtaining samples of liquid and gas from soil

    DOE Patents [OSTI]

    Rossabi, J.; May, C.P.; Pemberton, B.E.; Shinn, J.; Sprague, K.

    1999-03-30T23:59:59.000Z

    An apparatus and process for obtaining samples of liquid and gas from subsurface soil is provided having filter zone adjacent an external expander ring. The expander ring creates a void within the soil substrate which encourages the accumulation of soil-borne fluids. The fluids migrate along a pressure gradient through a plurality of filters before entering a first chamber. A one-way valve regulates the flow of fluid into a second chamber in further communication with a collection tube through which samples are collected at the surface. A second one-way valve having a reverse flow provides additional communication between the chambers for the pressurized cleaning and back-flushing of the apparatus. 8 figs.

  3. Process and apparatus for obtaining samples of liquid and gas from soil

    DOE Patents [OSTI]

    Rossabi, Joseph (105 Michael Ct., Aiken, SC 29801); May, Christopher P. (5002 Hesperus Dr., Columbia, MD 21044); Pemberton, Bradley E. (131 Glencarin Dr., Aiken, SC 29803); Shinn, Jim (Box 65, RFD. #1, South Royalton, VT 05068); Sprague, Keith (Box 234 Rte. 14, Brookfield, VT 05036)

    1999-01-01T23:59:59.000Z

    An apparatus and process for obtaining samples of liquid and gas from subsurface soil is provided having filter zone adjacent an external expander ring. The expander ring creates a void within the soil substrate which encourages the accumulation of soil-borne fluids. The fluids migrate along a pressure gradient through a plurality of filters before entering a first chamber. A one-way valve regulates the flow of fluid into a second chamber in further communication with a collection tube through which samples are collected at the surface. A second one-way valve having a reverse flow provides additional communication between the chambers for the pressurized cleaning and back-flushing of the apparatus.

  4. Precipitation-adsorption process for the decontamination of nuclear waste supernates

    DOE Patents [OSTI]

    Lee, Lien-Mow (North Augusta, SC); Kilpatrick, Lester L. (Aiken, SC)

    1984-01-01T23:59:59.000Z

    High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

  5. Precipitation-adsorption process for the decontamination of nuclear waste supernates

    DOE Patents [OSTI]

    Lee, L.M.; Kilpatrick, L.L.

    1982-05-19T23:59:59.000Z

    High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

  6. How to Put the Dollar Value on Waste Heat Recovery in the Process Industry 

    E-Print Network [OSTI]

    Campagne, W. V. L.

    1982-01-01T23:59:59.000Z

    Waste heat recovery projects should be evaluated on their actual fuel savings and not on Btu recovery. By equating waste heat recovery with potential steam savings, the fuel (or dollar) values of the waste heat as function of its temperature can...

  7. How to Put the Dollar Value on Waste Heat Recovery in the Process Industry

    E-Print Network [OSTI]

    Campagne, W. V. L.

    1982-01-01T23:59:59.000Z

    Waste heat recovery projects should be evaluated on their actual fuel savings and not on Btu recovery. By equating waste heat recovery with potential steam savings, the fuel (or dollar) values of the waste heat as function of its temperature can...

  8. Waste Receiving and Processing (WRAP) Facility Public Address System Review Findings

    SciTech Connect (OSTI)

    HUMPHRYS, K.L.

    1999-11-03T23:59:59.000Z

    Public address system operation at the Waste Receiving and Processing (WRAP) facility was reviewed. The review was based on an Operational Readiness Review finding that public address performance was not adequate in parts of the WRAP facility. Several improvements were made to the WRAP Public Address (PA) system to correct the deficiencies noted. Speaker gain and position was optimized. A speech processor was installed to boost intelligibility in high noise areas. Additional speakers were added to improve coverage in the work areas. The results of this evaluation indicate that further PA system enhancements are not warranted. Additional speakers cannot compensate for the high background sound and high reverberation levels found in the work areas. Recommendations to improve PA system intelligibility include minor speaker adjustments, enhanced PA announcement techniques, and the use of sound reduction and abatement techniques where economically feasible.

  9. Commercial-Scale Demonstration of the Liquid Phase Methanol (LPMEOTH) Process

    SciTech Connect (OSTI)

    None

    1998-12-21T23:59:59.000Z

    The Liquid Phase Methanol (LPMEOW) Demonstration Project at Kingsport, Tennessee, is a $213.7 million cooperative agreement between the U.S. Department of Energy (DOE) and Air Products Liquid Phase Conversion Company, L.P. (the Partnership) to produce methanol from coal-derived synthesis gas (syngas). Air Products and Chemicals, Inc. (Air Products) and Eastman Chemical Company (Eastman) formed the Partnership to execute the Demonstration Project. The LPMEOI-P Process Demonstration Unit was built at a site located at the Eastman coal-to-chemicals complex in Kingsport. During this quarter, initial planning and procurement work continued on the seven project sites which have been accepted for participation in the off-site, product-use test program. Approximately 12,000 gallons of fuel-grade methanol (98+ wt% methanol, 4 wt% water) produced during operation on carbon monoxide (CO)-rich syngas at the LPMEOW Demonstration Unit was loaded into trailers and shipped off-site for Mure product-use testing. At one of the projects, three buses have been tested on chemical-grade methanol and on fhel-grade methanol from the LPMEOW Demonstration Project. During the reporting period, planning for a proof-of-concept test run of the Liquid Phase Dimethyl Ether (LPDME~ Process at the Alternative Fuels Development Unit (AFDU) in LaPorte, TX continued. The commercial catalyst manufacturer (Calsicat) has prepared the first batch of dehydration catalyst in large-scale equipment. Air Products will test a sample of this material in the laboratory autoclave. Catalyst activity, as defined by the ratio of the rate constant at any point in time to the rate constant for freshly reduced catalyst (as determined in the laborato~ autoclave), was monitored for the initial extended operation at the lower initial reactor operating temperature of 235oC. At this condition, the decrease in catalyst activity with time from the period 20 December 1997 through 27 January 1998 occurred at a rate of 1.0% per day, which represented a significant improvement over the 3.4Yi per day decline measured during the initial six weeks of operation in April and May of 1997. The deactivation rate also improved from the longer-term rate of 1.6% per day calculated throughout the summer and autumn of 1997.

  10. Verification of the Accountability Method as a Means to Classify Radioactive Wastes Processed Using THOR Fluidized Bed Steam Reforming at the Studsvik Processing Facility in Erwin, Tennessee, USA - 13087

    SciTech Connect (OSTI)

    Olander, Jonathan [Studsvik Processing Facility Erwin, 151 T.C. Runnion Rd., Erwin, TN 37650 (United States)] [Studsvik Processing Facility Erwin, 151 T.C. Runnion Rd., Erwin, TN 37650 (United States); Myers, Corey [Studsvik, Inc., 5605 Glenridge Drive, Suite 705, Atlanta, GA 30342 (United States)] [Studsvik, Inc., 5605 Glenridge Drive, Suite 705, Atlanta, GA 30342 (United States)

    2013-07-01T23:59:59.000Z

    Studsviks' Processing Facility Erwin (SPFE) has been treating Low-Level Radioactive Waste using its patented THOR process for over 13 years. Studsvik has been mixing and processing wastes of the same waste classification but different chemical and isotopic characteristics for the full extent of this period as a general matter of operations. Studsvik utilizes the accountability method to track the movement of radionuclides from acceptance of waste, through processing, and finally in the classification of waste for disposal. Recently the NRC has proposed to revise the 1995 Branch Technical Position on Concentration Averaging and Encapsulation (1995 BTP on CA) with additional clarification (draft BTP on CA). The draft BTP on CA has paved the way for large scale blending of higher activity and lower activity waste to produce a single waste for the purpose of classification. With the onset of blending in the waste treatment industry, there is concern from the public and state regulators as to the robustness of the accountability method and the ability of processors to prevent the inclusion of hot spots in waste. To address these concerns and verify the accountability method as applied by the SPFE, as well as the SPFE's ability to control waste package classification, testing of actual waste packages was performed. Testing consisted of a comprehensive dose rate survey of a container of processed waste. Separately, the waste package was modeled chemically and radiologically. Comparing the observed and theoretical data demonstrated that actual dose rates were lower than, but consistent with, modeled dose rates. Moreover, the distribution of radioactivity confirms that the SPFE can produce a radiologically homogeneous waste form. The results of the study demonstrate: 1) the accountability method as applied by the SPFE is valid and produces expected results; 2) the SPFE can produce a radiologically homogeneous waste; and 3) the SPFE can effectively control the waste package classification. (authors)

  11. Spent Fuel and Waste Management Technology Development Program. Annual progress report

    SciTech Connect (OSTI)

    Bryant, J.W.

    1994-01-01T23:59:59.000Z

    This report provides information on the progress of activities during fiscal year 1993 in the Spent Fuel and Waste Management Technology Development Program (SF&WMTDP) at the Idaho Chemical Processing Plant (ICPP). As a new program, efforts are just getting underway toward addressing major issues related to the fuel and waste stored at the ICPP. The SF&WMTDP has the following principal objectives: Investigate direct dispositioning of spent fuel, striving for one acceptable waste form; determine the best treatment process(es) for liquid and calcine wastes to minimize the volume of high level radioactive waste (HLW) and low level waste (LLW); demonstrate the integrated operability and maintainability of selected treatment and immobilization processes; and assure that implementation of the selected waste treatment process is environmentally acceptable, ensures public and worker safety, and is economically feasible.

  12. CONTAMINATED PROCESS EQUIPMENT REMOVAL FOR THE D&D OF THE 232-Z CONTAMINATED WASTE RECOVERY PROCESS FACILITY AT THE PLUTONIUM FINISHING PLANT (PFP)

    SciTech Connect (OSTI)

    HOPKINS, A.M.; MINETTE, M.J.; KLOS, D.B.

    2007-01-25T23:59:59.000Z

    This paper describes the unique challenges encountered and subsequent resolutions to accomplish the deactivation and decontamination of a plutonium ash contaminated building. The 232-Z Contaminated Waste Recovery Process Facility at the Plutonium Finishing Plant was used to recover plutonium from process wastes such as rags, gloves, containers and other items by incinerating the items and dissolving the resulting ash. The incineration process resulted in a light-weight plutonium ash residue that was highly mobile in air. This light-weight ash coated the incinerator's process equipment, which included gloveboxes, blowers, filters, furnaces, ducts, and filter boxes. Significant airborne contamination (over 1 million derived air concentration hours [DAC]) was found in the scrubber cell of the facility. Over 1300 grams of plutonium held up in the process equipment and attached to the walls had to be removed, packaged and disposed. This ash had to be removed before demolition of the building could take place.

  13. Advance Digital Imaging Process for Tungsten Alloys Liquid-Phase Sintered in Microgravity W. B Goodwin, University of Tennessee, SURF 2009 Fellow

    E-Print Network [OSTI]

    Li, Mo

    Advance Digital Imaging Process for Tungsten Alloys Liquid-Phase Sintered in Microgravity W. B is to develop a digital image processing algorithm to rapidly detect and measure dihedral angles in binary. This geometric parameter greatly affects the liquid-phase sintering (LPS) process which is commonly used

  14. Process for removing and detoxifying cadmium from scrap metal including mixed waste

    SciTech Connect (OSTI)

    Kronberg, J.W.

    1994-07-01T23:59:59.000Z

    Cadmium-bearing scrap from nuclear applications, such as neutron shielding and reactor control and safety rods, must usually be handled as mixed waste since it is radioactive and the cadmium in it is both leachable and highly toxic. Removing the cadmium from this scrap, and converting it to a nonleachable and minimally radioactive form, would greatly simplify disposal or recycling. A process now under development will do this by shredding the scrap; leaching it with reagents which selectively dissolve out the cadmium; reprecipitating the cadmium as its highly insoluble sulfide; then fusing the sulfide into a glassy matrix to bring its leachability below EPA limits before disposal. Alternatively, the cadmium may be recovered for reuse. A particular advantage of the process is that all reagents (except the glass frit) can easily be recovered and reused in a nearly closed cycle, minimizing the risk of radioactive release. The process does not harm common metals such as aluminum, iron and stainless steel, and is also applicable to non-nuclear cadmium-bearing scrap such as nickel-cadmium batteries.

  15. A Low-cost, High-yield Process for the Direct Productin of High Energy Density Liquid Fuel from Biomass

    SciTech Connect (OSTI)

    Agrawal, Rakesh

    2014-02-21T23:59:59.000Z

    The primary objective and outcome of this project was the development and validation of a novel, low-cost, high-pressure fast-hydropyrolysis/hydrodeoxygenation (HDO) process (H{sub 2}Bioil) using supplementary hydrogen (H{sub 2}) to produce liquid hydrocarbons from biomass. The research efforts under the various tasks of the project have culminated in the first experimental demonstration of the H2Bioil process, producing 100% deoxygenated >C4+ hydrocarbons containing 36-40% of the carbon in the feed of pyrolysis products from biomass. The demonstrated H{sub 2}Bioil process technology (i.e. reactor, catalyst, and downstream product recovery) is scalable to a commercial level and is estimated to be economically competitive for the cases when supplementary H{sub 2} is sourced from coal, natural gas, or nuclear. Additionally, energy systems modeling has revealed several process integration options based on the H{sub 2}Bioil process for energy and carbon efficient liquid fuel production. All project tasks and milestones were completed or exceeded. Novel, commercially-scalable, high-pressure reactors for both fast-hydropyrolysis and hydrodeoxygenation were constructed, completing Task A. These reactors were capable of operation under a wide-range of conditions; enabling process studies that lead to identification of optimum process conditions. Model compounds representing biomass pyrolysis products were studied, completing Task B. These studies were critical in identifying and developing HDO catalysts to target specific oxygen functional groups. These process and model compound catalyst studies enabled identification of catalysts that achieved 100% deoxygenation of the real biomass feedstock, sorghum, to form hydrocarbons in high yields as part of Task C. The work completed during this grant has identified and validated the novel and commercially scalable H2Bioil process for production of hydrocarbon fuels from biomass. Studies on model compounds as well as real biomass feedstocks were utilized to identify optimized process conditions and selective HDO catalyst for high yield production of hydrocarbons from biomass. In addition to these experimental efforts, in Tasks D and E, we have developed a mathematical optimization framework to identify carbon and energy efficient biomass-to-liquid fuel process designs that integrate the use of different primary energy sources along with biomass (e.g. solar, coal or natural gas) for liquid fuel production. Using this tool, we have identified augmented biomass-to-liquid fuel configurations based on the fast-hydropyrolysis/HDO pathway, which was experimentally studied in this project. The computational approach used for screening alternative process configurations represents a unique contribution to the field of biomass processing for liquid fuel production.

  16. Cross flow filtration of aqueous radioactive tank wastes

    SciTech Connect (OSTI)

    McCabe, D.J. [Westinghouse Savannah River Co., Aiken, SC (United States); Reynolds, B.A. [Battelle Pacific Northwest Lab., Richland, WA (United States); Todd, T.A. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States); Wilson, J.H. [Oak Ridge National Lab., TN (United States)

    1997-02-01T23:59:59.000Z

    The Tank Focus Area (TFA) of the Department of Energy (DOE) Office of Science and Technology addresses remediation of radioactive waste currently stored in underground tanks. Baseline technologies for treatment of tank waste can be categorized into three types of solid liquid separation: (a) removal of radioactive species that have been absorbed or precipitated, (b) pretreatment, and (c) volume reduction of sludge and wash water. Solids formed from precipitation or absorption of radioactive ions require separation from the liquid phase to permit treatment of the liquid as Low Level Waste. This basic process is used for decontamination of tank waste at the Savannah River Site (SRS). Ion exchange of radioactive ions has been proposed for other tank wastes, requiring removal of insoluble solids to prevent bed fouling and downstream contamination. Additionally, volume reduction of washed sludge solids would reduce the tank space required for interim storage of High Level Wastes. The scope of this multi-site task is to evaluate the solid/liquid separations needed to permit treatment of tank wastes to accomplish these goals. Testing has emphasized cross now filtration with metal filters to pretreat tank wastes, due to tolerance of radiation and caustic.

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

    SciTech Connect (OSTI)

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

    2012-06-11T23:59:59.000Z

    During irradiation of nuclear fuel in a reactor, the five metals, Mo, Pd, Rh, Ru, and Tc, migrate to the fuel grain boundaries and form small metal particles of an alloy known as epsilon metal ({var_epsilon}-metal). When the fuel is dissolved in a reprocessing plant, these metal particles remain behind with a residue - the undissolved solids (UDS). Some of these same metals that comprise this alloy that have not formed the alloy are dissolved into the aqueous stream. These metals limit the waste loading for a borosilicate glass that is being developed for the reprocessing wastes. Epsilon metal is being developed as a waste form for the noble metals from a number of waste streams in the aqueous reprocessing of used nuclear fuel (UNF) - (1) the {var_epsilon}-metal from the UDS, (2) soluble Tc (ion-exchanged), and (3) soluble noble metals (TRUEX raffinate). Separate immobilization of these metals has benefits other than allowing an increase in the glass waste loading. These materials are quite resistant to dissolution (corrosion) as evidenced by the fact that they survive the chemically aggressive conditions in the fuel dissolver. Remnants of {var_epsilon}-metal particles have survived in the geologically natural reactors found in Gabon, Africa, indicating that they have sufficient durability to survive for {approx} 2.5 billion years in a reducing geologic environment. Additionally, the {var_epsilon}-metal can be made without additives and incorporate sufficient foreign material (oxides) that are also present in the UDS. Although {var_epsilon}-metal is found in fuel and Gabon as small particles ({approx}10 {micro}m in diameter) and has survived intact, an ideal waste form is one in which the surface area is minimized. Therefore, the main effort in developing {var_epsilon}-metal as a waste form is to develop a process to consolidate the particles into a monolith. Individually, these metals have high melting points (2617 C for Mo to 1552 C for Pd) and the alloy is expected to have a high melting point as well, perhaps exceeding 1500 C. The purpose of the work reported here is to find a potential commercial process with which {var_epsilon}-metal plus other components of UDS can be consolidated into a solid with minimum surface area and high strength Here, we report the results from the preliminary evaluation of spark-plasma sintering (SPS), hot-isostatic pressing (HIP), and microwave sintering (MS). Since bulk {var_epsilon}-metal is not available and companies could not handle radioactive materials, we prepared mixtures of the five individual metal powders (Mo, Ru, Rh, Pd, and Re) and baddeleyite (ZrO{sub 2}) to send the vendors of SPS, HIP, and MS. The processed samples were then evaluated at the Pacific Northwest National Laboratory (PNNL) for bulk density and phase assemblage with X-ray diffraction (XRD) and phase composition with scanning electron microscopy (SEM). Physical strength was evaluated qualitatively. Results of these scoping tests showed that fully dense cermet (ceramic-metal composite) materials with up to 35 mass% of ZrO{sub 2} were produced with SPS and HIP. Bulk density of the SPS samples ranged from 87 to 98% of theoretical density, while HIP samples ranged from 96 to 100% of theoretical density. Microwave sintered samples containing ZrO{sub 2} had low densities of 55 to 60% of theoretical density. Structurally, the cermet samples showed that the individual metals alloyed in to {var_epsilon}-phase - hexagonal-close-packed (HCP) alloy (4-95 mass %), the {alpha}-phase - face-centered-cubic (FCC) alloy structure (3-86 mass %), while ZrO{sub 2} remained in the monoclinic structure of baddeleyite. Elementally, the samples appeared to have nearly uniform composition, but with some areas rich in Mo and Re, the two components with the highest melting points. The homogeneity in distribution of the elements in the alloy is significantly improved in the presence of ZrO{sub 2}. However, ZrO{sub 2} does not appear to react with the alloy, nor was Zr found in the alloy.

  18. Low Emissions Burner Technology for Metal Processing Industry using Byproducts and Biomass Derived Liquid Fuels

    SciTech Connect (OSTI)

    Agrawal, Ajay; Taylor, Robert

    2013-09-30T23:59:59.000Z

    This research and development efforts produced low-emission burner technology capable of operating on natural gas as well as crude glycerin and/or fatty acids generated in biodiesel plants. The research was conducted in three stages (1) Concept definition leading to the design and development of a small laboratory scale burner, (2) Scale-up to prototype burner design and development, and (3) Technology demonstration with field vefiication. The burner design relies upon the Flow Blurring (FB) fuel injection based on aerodynamically creating two-phase flow near the injector exit. The fuel tube and discharge orifice both of inside diameter D are separated by gap H. For H < 0.25D, the atomizing air bubbles into liquid fuel to create a two-phase flow near the tip of the fuel tube. Pressurized two-phase fuel-air mixture exits through the discharge orifice, which results in expansion and breakup of air bubbles yielding a spray with fine droplets. First, low-emission combustion of diesel, biodiesel and straight VO (soybean oil) was achieved by utilizing FB injector to yield fine sprays for these fuels with significantly different physical properties. Visual images for these baseline experiments conducted with heat release rate (HRR) of about 8 kW illustrate clean blue flames indicating premixed combustion for all three fuels. Radial profiles of the product gas temperature at the combustor exit overlap each other signifying that the combustion efficiency is independent of the fuel. At the combustor exit, the NOx emissions are within the measurement uncertainties, while CO emissions are slightly higher for straight VO as compared to diesel and biodiesel. Considering the large variations in physical and chemical properties of fuels considered, the small differences observed in CO and NOx emissions show promise for fuel-flexible, clean combustion systems. FB injector has proven to be very effective in atomizing fuels with very different physical properties, and it offers a path forward to utilize both fossil and alternative liquid fuels in the same combustion system. In particular, experiments show that straight VO can be cleanly combusted without the need for chemical processing or preheating steps, which can result in significant economic and environmental benefits. Next, low-emission combustion of glycerol/methane was achieved by utilizing FB injector to yield fine droplets of highly viscous glycerol. Heat released from methane combustion further improves glycerol pre-vaporization and thus its clean combustion. Methane addition results in an intensified reaction zone with locally high temperatures near the injector exit. Reduction in methane flow rate elongates the reaction zone, which leads to higher CO emissions and lower NOx emissions. Similarly, higher air to liquid (ALR) mass ratio improves atomization and fuel pre-vaporization and shifts the flame closer to the injector exit. In spite of these internal variations, all fuel mixes of glycerol with methane produced similar CO and NOx emissions at the combustor exit. Results show that FB concept provides low emissions with the flexibility to utilize gaseous and highly viscous liquid fuels, straight VO and glycerol, without preheating or preprocessing the fuels. Following these initial experiments in quartz combustor, we demonstrated that glycerol combustion can be stably sustained in a metal combustor. Phase Doppler Particle Analyzer (PDPA) measurements in glycerol/methane flames resulted in flow-weighted Sauter Mean Diameter (SMD) of 35 to 40 ?m, depending upon the methane percentage. This study verified that lab-scale dual-fuel burner using FB injector can successfully atomize and combust glycerol and presumably other highly viscous liquid fuels at relatively low HRR (<10 kW). For industrial applications, a scaled-up glycerol burner design thus seemed feasible.

  19. A low-temperature process for the denitration of Hanford single-shell tank, nitrate-based waste utilizing the nitrate to ammonia and ceramic (NAC) process

    SciTech Connect (OSTI)

    Mattus, A.J.; Lee, D.D.; Dillow, T.A.; Farr, L.L.; Loghry, S.L.; Pitt, W.W.; Gibson, M.R.

    1994-12-01T23:59:59.000Z

    Bench-top feasibility studies with Hanford single-shell tank (SST) simulants, using a new, low-temperature (50 to 60C) process for converting nitrate to ammonia and ceramic (NAC), have conclusively shown that between 85 to 99% of the nitrate can be readily converted. In this process, aluminum powders or shot can be used to convert alkaline, nitrate-based supernate to ammonia and an aluminum oxide-sodium aluminate-based solid which might function as its own waste form. The process may actually be able to utilize already contaminated aluminum scrap metal from various DOE sites to effect the conversion. The final, nearly nitrate-free ceramic-like product can be pressed and sintered like other ceramics. Based upon the starting volumes of 6.2 and 3.1 M sodium nitrate solution, volume reductions of 50 to 55% were obtained for the waste form produced, compared to an expected 35 to 50% volume increase if the Hanford supernate were grouted. Engineering data extracted from bench-top studies indicate that the process will be very economical to operate, and data were used to cost a batch, 1,200-kg NO{sub 3}/h plant for working off Hanford SST waste over 20 years. Their total process cost analysis presented in the appendix, indicates that between $2.01 to 2.66 per kilogram of nitrate converted will be required. Additionally, data on the fate of select radioelements present in solution are presented in this report as well as kinetic, operational, and control data for a number of experiments. Additionally, if the ceramic product functions as its own waste form, it too will offer other cost savings associated with having a smaller volume of waste form as well as eliminating other process steps such as grouting.

  20. Field study of disposed solid wastes from advanced coal processes. Annual technical progress report, October 1991--September 1992

    SciTech Connect (OSTI)

    Not Available

    1992-12-31T23:59:59.000Z

    Radian Corporation and the North Dakota Energy and Environmental Research Center (EERC) are funded to develop information to be used by private industry and government agencies for managing solid wastes produced by advanced coal combustion processes. This information will be developed by conducting several field studies on disposed wastes from these processes. Data will be collected to characterize these wastes and their interactions with the environments in which they are disposed. Three sites were selected for the field studies: Colorado Ute`s fluidized bed combustion (FBC) unit in Nucla, Colorado; Ohio Edison`s limestone injection multistage burner (LIMB) retrofit in Lorain, Ohio; and Freeman United`s mine site in central Illinois with wastes supplied by the nearby Midwest Grain FBC unit. During the past year, field monitoring and sampling of the four landfill test cases constructed in 1989 and 1991 has continued. Option 1 of the contract was approved last year to add financing for the fifth test case at the Freeman United site. The construction of the Test Case 5 cells is scheduled to begin in November, 1992. Work during this past year has focused on obtaining data on the physical and chemical properties of the landfilled wastes, and on developing a conceptual framework for interpreting this information. Results to date indicate that hydration reactions within the landfilled wastes have had a major impact on the physical and chemical properties of the materials but these reactions largely ceased after the first year, and physical properties have changed little since then. Conditions in Colorado remained dry and no porewater samples were collected. In Ohio, hydration reactions and increases in the moisture content of the waste tied up much of the water initially infiltrating the test cells.