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Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


1

Newly Generated Liquid Waste Processing Alternatives Study, Volume 1  

SciTech Connect

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.

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

2002-09-01T23:59:59.000Z

2

Liquid Waste Processing Facilities (LWPF) Reliability and Availability and Maintainability (RAM) Analysis  

SciTech Connect

A reliability, availability, and maintainability (RAM) analysis was prepared for the liquid effluents support being provided to the River Protection Project Waste Treatment Plant (WTP). The availability of liquid effluents services to the WTP was determined. Recommendations are provided on improvements and upgrades to increase the availability of the Liquid Waste Processing Facilities treatment and disposal systems.

LOWE, S.S.

2001-02-20T23:59:59.000Z

3

An evaluation of neutralization for processing sodium-bearing liquid waste  

SciTech Connect

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.

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

1989-01-01T23:59:59.000Z

4

Process for immobilizing radioactive boric acid liquid wastes  

DOE Patents (OSTI)

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.

Greenhalgh, W.O.

1984-05-10T23:59:59.000Z

5

Process for immobilizing radioactive boric acid liquid wastes  

DOE Patents (OSTI)

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.

Greenhalgh, Wilbur O. (Richland, WA)

1986-01-01T23:59:59.000Z

6

Use of Chemical Pretreatment to Enhance Liquid Waste of Processing  

Science Conference Proceedings (OSTI)

This report details unique chemical pretreatment and processing techniques presently used by utilities in their nuclear plant liquid radwaste treatment programs. It presents specific utility experience from a number of plants employing these processing alternatives.

1999-04-23T23:59:59.000Z

7

EA-437; Environmental Assessment Process Equipment Waste and Process Waste Liquid Collection Systems Idaho Chemical Processing Plant Idaho National Engineering Laboratory  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

437; Environmental Assessment Process Equipment Waste and 437; Environmental Assessment Process Equipment Waste and Process Waste Liquid Collection Systems Idaho Chemical Processing Plant Idaho National Engineering Laboratory TABLE OF CONTENTS Environmental Assessment Process Equipment Waste and Process Waste Liquid Collection Systems Idaho Chemical Processing Plant Idaho National Engineering Laboratory 1. INTRODUCTION 2. DESCRIPTION OF THE PROPOSED ACTION AND ALTERNATIVES 2.1 Purpose and Need of the Proposed Action 2.2 Description of the Affected Facilities 2.3 Description of Proposed Action 2.4 Alternatives to the Proposed Action 2.5 Separate But Related Actions 3. AFFECTED ENVIRONMENT 3.1 Introduction 3.2 Physical Environment 3.3 Biological Resources 3.4 Cultural Resources 3.5 Environmental Quality and Monitoring Programs

8

Process for Removing Radioactive Wastes from Liquid Streams  

SciTech Connect

The process is under development at Mound Laboratory to remove radioactive waste (principally plutonium-238) from process water prior to discharge of the water to the Miami river. The contaminated water, as normally received, is at a pH between 6 and 90. Under these conditions, plutonium in all its oxidation states is hydrolyzed; however, the level of the radioactive solids varies from about 50ppm down to about 50 ppb and the plutonium remains in a colloidal or subcolloidal condition. The permissible concentration for discharge to the river is about 50 parts per trillion. Pilot plant test show that 95-99% of the radioactive material is removed by adsorption on diatomaceous earth. The remainder is removed by passage through a bed of either dibasic or tribasic calcium phosphate. Ground phosphate rock is equally effective in removing the radioactive material if the flow rate is controlled to permit sufficient contact time. Parameters for optimizing the process are now under study. Future plans include application of the process to wastes from reactor fuels reprocessing.

Kirby, H. W.; Blane, D. E.; Smolin, R. I.

1972-10-01T23:59:59.000Z

9

Evaluation of System Level Modeling and Simulation Tools in Support of Savannah River Site Liquid Waste Process  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Savannah River Site Liquid Waste Process Savannah River Site Liquid Waste Process June 2009 Monica C. Regalbuto Office of Waste Processing DOE/EM Kevin G. Brown Vanderbilt University and CRESP David W. DePaoli Oak Ridge National Laboratory Candido Pereira Argonne National Laboratory John R. Shultz Office of Waste Processing DOE/EM Sahid C. Smith Office of Waste Processing DOE/EM External Technical Review for Evaluation of System Level Modeling and Simulation Tools in Support of Savannah River Site Liquid Waste Process June 2009 ACKNOWLEDGEMENTS The Review Team thanks Ms. Sonitza Blanco, Team Lead Planning and Coordination Waste Disposition Project U.S. Department of Energy Savannah River Operations Office and Mr. Pete Hill, Liquid Waste Planning Manager for Washington Savannah River Company, for their

10

External Technical Review for Evaluation of System Level Modeling and Simulation Tools in Support of Hanford Site Liquid Waste Process  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Hanford Site Liquid Waste Process Hanford Site Liquid Waste Process September 2009 Monica C. Regalbuto Office of Waste Processing DOE/EM Kevin G. Brown Vanderbilt University and CRESP David W. DePaoli Oak Ridge National Laboratory Candido Pereira Argonne National Laboratory John R. Shultz Office of Waste Processing DOE/EM External Technical Review for Evaluation of System Level Modeling and Simulation Tools in Support of Hanford Site Liquid Waste Process September 2009 Acknowledgements The Review Team thanks Mr. Glyn Trenchard, Team Lead for Planning and Coordination Waste Disposition Project, U.S. Department of Energy--Office of River Protection, Mr. Paul Rutland, RPP System Planning Manager for Washington River Protection Solutions, and Mr. Ernie Lee,

11

SRS - Programs - Liquid Waste Disposition  

NLE Websites -- All DOE Office Websites (Extended Search)

Liquid Waste Disposition Liquid Waste Disposition This includes both the solidification of highly radioactive liquid wastes stored in SRS's tank farms and disposal of liquid low-level waste generated as a by-product of the separations process and tank farm operations. This low-level waste is treated in the Effluent Treatment Facility. High-activity liquid waste is generated at SRS as by-products from the processing of nuclear materials for national defense, research and medical programs. The waste, totaling about 36 million gallons, is currently stored in 49 underground carbon-steel waste tanks grouped into two "tank farms" at SRS. While the waste is stored in the tanks, it separates into two parts: a sludge that settles on the bottom of the tank, and a liquid supernate that resides on top of the sludge. The waste is reduced to about 30 percent of its original volume by evaporation. The condensed evaporator "overheads" are transferred to the Effluent Treatment Project for final cleanup prior to release to the environment. As the concentrate cools a portion of it crystallizes forming solid saltcake. The concentrated supernate and saltcake are less mobile and therefore less likely to escape to the environment in the event of a tank crack or leak.

12

MUSHROOM WASTE MANAGEMENT PROJECT LIQUID WASTE MANAGEMENT  

E-Print Network (OSTI)

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

13

Radioactive Liquid Processing Guidelines  

Science Conference Proceedings (OSTI)

This report presents guidance for utility liquid radwaste processing program managers. The document is a summation of utility and vendor processing experience, and is intended for use as a tool to enhance liquid radwaste processing programs. Utilization of this information will result in optimized system performance, and a reduction in waste volumes and program costs.

2005-11-22T23:59:59.000Z

14

Development of the SREX process for the treatment of ICPP liquid wastes  

SciTech Connect

The removal of {sup 90}Sr from actual and simulated wastes at the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering and Environmental Laboratory (INEEL) has been demonstrated with the SREX process. This solvent extraction process employs the extractant 4{prime},4{prime}(5{prime}) di-(t-butylcyclohexano)-18-crown-6 in 1-octanol or a mixture of tributyl phosphate and a hydrocarbon diluent called Isopar L{reg_sign}. Process flowsheets have been designed for testing in countercurrent experiments with centrifugal contractors. The flowsheets have been designed using batch contract solvent extraction methods. The extraction of Sr as well as other interfering ions has been studied. The effect of various parameters including nitric acid dependence, extractant concentration dependence, hydronium ion concentration, and interferent concentrations upon the extraction efficiency of the process has been evaluated. The radiolysis of the SREX solvent has also been investigated as a function of absorbed gamma radiation. The extraction efficiency of the solvent has been shown to be only slightly dependent upon absorbed dose in the range 0--1,000 kGy. The decontamination of actual sodium-bearing waste and dissolved calcine solutions has been accomplished in batch contact flowsheets. Decontamination factors as high as 10E3 have been obtained with sequential batch contacts. Flowsheets have been developed to accomplish decontamination of the liquid wastes with respect to {sup 90}Sr as well as the removal of Pb and Hg. Pb may be partitioned from the Sr fraction in a separate stripping procedure using ammonium citrate. This work has led to the formulation of countercurrent flowsheets which have been tested in centrifugal contractors with actual waste and reported in the document INEEL/EXT-97-00832.

Wood, D.J.; Law, J.D.; Garn, T.G.; Tillotson, R.D.; Tullock, P.A.; Todd, T.A.

1997-12-01T23:59:59.000Z

15

Development of the SREX Process for the Treatment of ICPP Liquid Wastes  

SciTech Connect

The removal of Sr-90 from actual and simulated wastes at the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering and Environmental Laboratory (INEEL) has been demonstrated with the SREX process. This solvent extraction process employs the extractant 4',4' (5') de-(t-butylcyclohexano)-18-crown-6 in 1-octanol or a mixture of tributyl phosphate and a hydrocarbon diluent called Isopar L. This development work is based upon earlier work performed by Horwitz, et al. at Argonne National Laboratory. Process flowsheets have been designed for testing in countercurrent experiments with centrifugal contactors. The flowsheets have been designed using batch contact solvent extraction methods. The extraction of Sr as well as other interfering ions has been studied. The effect of various parameters including nitric acid dependence, extractant concentration dependence, Hydronium ion concentration, and interferent concentrations upon the extraction efficiency of the process has been evaluated. The radiolysis of the SREX solvent has also been investigated as a function of absorbed gamma radiation. The extraction efficiency of the solvent has been shown to be only slightly dependent upon absorbed dose in the range 0-1000 kGy. The decontamination of actual sodium-bearing waste and dissolved calcine solutions has been accomplished in batch contact flowsheets. Decontamination factors as high as 10E3 have been obtained with sequential batch contacts. Flowsheets have been developed to accomplish decontamination of the liquid wastes with respect to Sr-90, as well as the removal of Pb and Hg. Pb may be partitioned from the Sr fraction in a separate stripping procedure using ammonium citrate. This work has led to the formulation of countercurrent flowsheets which have been tested in centrifugal contactors with actual waste and reported in the document INEEL/EXT-97-00832.

D. J. Wood; Garn, T. G.; J. D. Law; P. A. Tullock; R. D. Tillotson; T. A. Todd

1997-10-01T23:59:59.000Z

16

Salt Waste Processing Initiatives  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Patricia Suggs Patricia Suggs Salt Processing Team Lead Assistant Manager for Waste Disposition Project Office of Environmental Management Savannah River Site Salt Waste Processing Initiatives 2 Overview * Current SRS Liquid Waste System status * Opportunity to accelerate salt processing - transformational technologies - Rotary Microfiltration (RMF) and Small Column Ion Exchange (SCIX) - Actinide Removal Process/Modular Caustic Side Solvent Extraction (ARP/MCU) extension with next generation extractant - Salt Waste Processing Facility (SWPF) performance enhancement - Saltstone enhancements * Life-cycle impacts and benefits 3 SRS Liquid Waste Total Volume >37 Million Gallons (Mgal) Total Curies 183 MCi (51% ) 175 MCi (49% ) >358 Million Curies (MCi) Sludge 34.3 Mgal (92% ) 3.0 Mgal (8%)

17

Waste Logic™ Liquid Waste Manager (WL-LWM) Software, Version 2.0  

Science Conference Proceedings (OSTI)

In response to continuing industry efforts to reduce operating expenditures, EPRI developed the Waste Logic&trade: Liquid Waste Manager code to analyze costs associated with liquid waste processing and the disposition of its resultant solid waste. EPRI's Waste Logic: Liquid Waste Manager software for windows-based PC computers provides a detailed economic and performance view of liquid waste processing activities. The software will help nuclear utilities evaluate the costs associated with liquid radwaste...

2002-06-05T23:59:59.000Z

18

Liquid-Liquid Extraction Processes  

E-Print Network (OSTI)

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 the two liquid phases, separation will result. This is the principle upon which separation by liquid-liquid extraction is based, and there are a number of important applications of this concept in industrial processes. This paper will review the basic concepts and applications as well as present future directions for the liquid-liquid extraction process.

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

1983-01-01T23:59:59.000Z

19

Process for converting sodium nitrate-containing, caustic liquid radioactive wastes to solid insoluble products  

DOE Patents (OSTI)

A method for converting sodium nitrate-containing, caustic, radioactive wastes to a solid, relatively insoluble, thermally stable form is provided and comprises the steps of reacting powdered aluminum silicate clay, e.g., kaolin, bentonite, dickite, halloysite, pyrophyllite, etc., with the sodium nitrate-containing radioactive wastes which have a caustic concentration of about 3 to 7 M at a temperature of 30.degree. C to 100.degree. C to thereby entrap the dissolved radioactive salts in the aluminosilicate matrix. In one embodiment the sodium nitrate-containing, caustic, radioactive liquid waste, such as neutralized Purex-type waste, or salts or oxide produced by evaporation or calcination of these liquid wastes (e.g., anhydrous salt cake) is converted at a temperature within the range of 30.degree. C to 100.degree. C to the solid mineral form-cancrinite having an approximate chemical formula 2(NaAlSiO.sub.4) .sup.. xSalt.sup.. y H.sub.2 O with x = 0.52 and y = 0.68 when the entrapped salt is NaNO.sub.3. In another embodiment the sodium nitrate-containing, caustic, radioactive liquid is reacted with the powdered aluminum silicate clay at a temperature within the range of 30.degree. C to 100.degree. C, the resulting reaction product is air dried eitheras loose powder or molded shapes (e.g., bricks) and then fired at a temperature of at least 600.degree. C to form the solid mineral form-nepheline which has the approximate chemical formula of NaAlSiO.sub.4. The leach rate of the entrapped radioactive salts with distilled water is reduced essentially to that of the aluminosilicate lattice which is very low, e.g., in the range of 10.sup.-.sup.2 to 10.sup.-.sup.4 g/cm.sup.2 -- day for cancrinite and 10.sup.-.sup.3 to 10.sup.-.sup.5 g/cm.sup.2 -- day for nepheline.

Barney, Gary S. (Richland, WA); Brownell, Lloyd E. (Richland, WA)

1977-01-01T23:59:59.000Z

20

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

E-Print Network (OSTI)

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

Russell, Lynn

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

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

E-Print Network (OSTI)

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

Firtel, Richard A.

22

Liquid low level waste management expert system  

SciTech Connect

An expert system has been developed as part of a new initiative for the Oak Ridge National Laboratory (ORNL) systems analysis program. This expert system will aid in prioritizing radioactive waste streams for treatment and disposal by evaluating the severity and treatability of the problem, as well as the final waste form. The objectives of the expert system development included: (1) collecting information on process treatment technologies for liquid low-level waste (LLLW) that can be incorporated in the knowledge base of the expert system, and (2) producing a prototype that suggests processes and disposal technologies for the ORNL LLLW system. 4 refs., 9 figs.

Ferrada, J.J.; Abraham, T.J. (Oak Ridge National Lab., TN (United States)); Jackson, J.R. (Southwest Baptist Univ., Bolivar, MO (USA))

1991-01-01T23:59:59.000Z

23

Method for treating liquid wastes  

DOE Patents (OSTI)

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.

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

24

Method for treating liquid wastes  

DOE Patents (OSTI)

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.

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

1995-12-26T23:59:59.000Z

25

Resource-Limited Multiattribute Value Analysis of Alternatives for Immobilizing Radioactive Liquid Process Waste Stored in Saluggia, Italy  

Science Conference Proceedings (OSTI)

This large Italian public works project started with the development of engineering data to support the evaluation of three alternatives for processing nuclear waste. After an analysis of the alternatives' performance from an engineering perspective ... Keywords: EUREX, Sogin, alternatives, applications, decision analysis, decision making, decision theory, energy, environment, multiattribute value analysis, nuclear waste

Alan J. Brothers; Shas V. Mattigod; Denis M. Strachan; Gordon H. Beeman; Paul K. Kearns; Angelo Papa; Carlo Monti

2009-06-01T23:59:59.000Z

26

Municipal waste processing apparatus  

DOE Patents (OSTI)

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.

Mayberry, J.L.

1988-04-13T23:59:59.000Z

27

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

Science Conference Proceedings (OSTI)

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.

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

2010-01-27T23:59:59.000Z

28

High-Level Liquid Waste Tank Integrity Workshop - 2008  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Liquid Waste Tank Integrity Liquid Waste Tank Integrity Workshop - 2008 Karthik Subramanian Bruce Wiersma November 2008 High Level Waste Corporate Board Meeting karthik.subramanian@srnl.doe.gov bruce.wiersma@srnl.doe.gov 2 Acknowledgements * Bruce Wiersma (SRNL) * Kayle Boomer (Hanford) * Michael T. Terry (Facilitator) * SRS - Liquid Waste Organization * Hanford Tank Farms * DOE-EM 3 Background * High level radioactive waste (HLW) tanks provide critical interim confinement for waste prior to processing and permanent disposal * Maintaining structural integrity (SI) of the tanks is a critical component of operations 4 Tank Integrity Workshop - 2008 * Discuss the HLW tank integrity technology needs based upon the evolving waste processing and tank closure requirements along with its continued storage mission

29

Salt Waste Processing Facility Fact Sheet | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Services » Waste Management » Tank Waste and Waste Processing » Services » Waste Management » Tank Waste and Waste Processing » Salt Waste Processing Facility Fact Sheet Salt Waste Processing Facility Fact Sheet Nuclear material production operations at SRS resulted in the generation of liquid radioactive waste that is being stored, on an interim basis, in 49 underground waste storage tanks in the F- and H-Area Tank Farms. SWPF Fact Sheet More Documents & Publications EIS-0082-S2: Amended Record of Decision Savannah River Site Salt Waste Processing Facility Technology Readiness Assessment Report EIS-0082-S2: Record of Decision Waste Management Nuclear Materials & Waste Tank Waste and Waste Processing Waste Disposition Packaging and Transportation Site & Facility Restoration Deactivation & Decommissioning (D&D)

30

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

SciTech Connect

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.

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

1994-02-01T23:59:59.000Z

31

Conversion of cellulosic wastes to liquid fuels  

DOE Green Energy (OSTI)

The current status and future plans for a project to convert waste cellulosic (biomass) materials to quality liquid hydrocarbon fuels is described. The basic approach is indirect liquefaction, i.e., thermal gasification followed by catalytic liquefaction. The indirect approach results in separation of the oxygen in the biomass feedstock, i.e., oxygenated compounds do not appear in the liquid hydrocarbon fuel product. The process is capable of accepting a wide variety of feedstocks. Potential products include medium quality gas, normal propanol, diesel fuel and/or high octane gasoline. A fluidized bed pyrolysis system is used for gasification. The pyrolyzer can be fluidized with recycle pyrolysis gas, steam or recycle liquefaction system off gas or some combination thereof. Tars are removed in a wet scrubber. Unseparated pyrolysis gases are utilized as feed to a modified Fischer-Tropsch reactor. The liquid condensate from the reactor consists of a normal propanol-water phase and a paraffinic hydrocarbon phase. The reactor can be operated to optimize for either product. The following tasks were specified in the statement of work for the contract period: (1) feedstock studies; (2) gasification system optimization; (3) waste stream characterization; and (4) liquid fuels synthesis. In addition, several equipment improvements were implemented.

Kuester, J.L.

1980-09-01T23:59:59.000Z

32

Liquid and Gaseous Waste Operations Department annual operating report CY 1996  

SciTech Connect

This annual report summarizes operating activities dealing with the process waste system, the liquid low-level waste system, and the gaseous waste system. It also describes upgrade activities dealing with the process and liquid low-level waste systems, the cathodic protection system, a stack ventilation system, and configuration control. Maintenance activities are described dealing with nonradiological wastewater treatment plant, process waste treatment plant and collection system, liquid low-level waste system, and gaseous waste system. Miscellaneous activities include training, audits/reviews/tours, and environmental restoration support.

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

1997-03-01T23:59:59.000Z

33

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

E-Print Network (OSTI)

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 ...

Hunt, Andrew Robert

2007-01-01T23:59:59.000Z

34

PROCESSING OF RADIOACTIVE WASTE  

DOE Patents (OSTI)

A process for treating radioactive waste solutions prior to disposal is described. A water-soluble phosphate, borate, and/or silicate is added. The solution is sprayed with steam into a space heated from 325 to 400 deg C whereby a powder is formed. The powder is melted and calcined at from 800 to 1000 deg C. Water vapor and gaseous products are separated from the glass formed. (AEC)

Johnson, B.M. Jr.; Barton, G.B.

1961-11-14T23:59:59.000Z

35

Recovery of Mercury From Contaminated Liquid Wastes  

SciTech Connect

The Base Contract program emphasized the manufacture and testing of superior sorbents for mercury removal, testing of the sorption process at a DOE site, and determination of the regeneration conditions in the laboratory. During this project, ADA Technologies, Inc. demonstrated the following key elements of a successful regenerable mercury sorption process: (1) sorbents that have a high capacity for dissolved, ionic mercury; (2) removal of ionic mercury at greater than 99% efficiency; and (3) thermal regeneration of the spent sorbent. ADA's process is based on the highly efficient and selective sorption of mercury by noble metals. Contaminated liquid flows through two packed columns that contain microporous sorbent particles on which a noble metal has been finely dispersed. A third column is held in reserve. When the sorbent is loaded with mercury to the point of breakthrough at the outlet of the second column, the first column is taken off-line and the flow of contaminated liquid is switched to the second and third columns. The spent column is regenerated by heating. A small flow of purge gas carries the desorbed mercury to a capture unit where the liquid mercury is recovered. Laboratory-scale tests with mercuric chloride solutions demonstrated the sorbents' ability to remove mercury from contaminated wastewater. Isotherms on surrogate wastes from DOE's Y-12 Plant in Oak Ridge, Tennessee showed greater than 99.9% mercury removal. Laboratory- and pilot-scale tests on actual Y-12 Plant wastes were also successful. Mercury concentrations were reduced to less than 1 ppt from a starting concentration of 1,000 ppt. The treatment objective was 50 ppt. The sorption unit showed 10 ppt discharge after six months. Laboratory-scale tests demonstrated the feasibility of sorbent regeneration. Results show that sorption behavior is not affected after four cycles.

1998-06-12T23:59:59.000Z

36

SRS Liquid Waste Program Partnering Agreement  

Energy.gov (U.S. Department of Energy (DOE))

We the members of the  SRS Liquid Waste Partnering Team do hereby mutually agree to work in a collaborative and cooperative manner through open communication and coordination with team members, and...

37

Method for solidifying liquid radioactive wastes  

DOE Patents (OSTI)

The quantity of nitrous oxides produced during the solidification of liquid radioactive wastes containing nitrates and nitrites can be substantially reduced by the addition to the wastes of a stoichiometric amount of urea which, upon heating, destroys the nitrates and nitrites, liberating nontoxic N.sub.2, CO.sub.2 and NH.sub.3.

Berreth, Julius R. (Idaho Falls, ID)

1976-01-01T23:59:59.000Z

38

ICPP radioactive liquid and calcine waste technologies evaluation. Interim report  

SciTech Connect

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.

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

1994-06-01T23:59:59.000Z

39

Radioactive waste processing apparatus  

DOE Patents (OSTI)

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.

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

40

PROCESSING OF RADIOACTIVE WASTE  

DOE Patents (OSTI)

A process for concentrating fission-product-containing waste solutions from fuel element processing is described. The process comprises the addition of sugar to the solution, preferably after it is made alkaline; spraying the solution into a heated space whereby a dry powder is formed; heating the powder to at least 220 deg C in the presence of oxygen whereby the powder ignites, the sugar is converted to carbon, and the salts are decomposed by the carbon; melting the powder at between 800 and 900 deg C; and cooling the melt. (AEC) antidiuretic hormone from the blood by the liver. Data are summarized from the following: tracer studies on cardiovascular functions; the determination of serum protein-bound iodine; urinary estrogen excretion in patients with arvanced metastatic mammary carcinoma; the relationship between alheroclerosis aad lipoproteins; the physical chemistry of lipoproteins; and factors that modify the effects of densely ionizing radia

Allemann, R.T.; Johnson, B.M. Jr.

1961-10-31T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Municipal waste processing apparatus  

DOE Patents (OSTI)

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

Mayberry, J.L.

1987-01-15T23:59:59.000Z

42

Municipal waste processing apparatus  

DOE Patents (OSTI)

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

Mayberry, John L. (Idaho Falls, ID)

1988-01-01T23:59:59.000Z

43

Municipal waste processing apparatus  

DOE Patents (OSTI)

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

Mayberry, John L. (Idaho Falls, ID)

1989-01-01T23:59:59.000Z

44

Radioactive waste processing apparatus  

DOE Patents (OSTI)

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.

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

1985-08-30T23:59:59.000Z

45

INEEL Radioactive Liquid Waste Reduction Program  

SciTech Connect

Reduction of radioactive liquid waste, much of which is Resource Conservation and Recovery Act (RCRA) listed, is a high priority at the Idaho National Technology and Engineering Center (INTEC). Major strides in the past five years have lead to significant decreases in generation and subsequent reduction in the overall cost of treatment of these wastes. In 1992, the INTEC, which is part of the Idaho National Environmental and Engineering Laboratory (INEEL), began a program to reduce the generation of radioactive liquid waste (both hazardous and non-hazardous). As part of this program, a Waste Minimization Plan was developed that detailed the various contributing waste streams, and identified methods to eliminate or reduce these waste streams. Reduction goals, which will reduce expected waste generation by 43%, were set for five years as part of this plan. The approval of the plan led to a Waste Minimization Incentive being put in place between the Department of Energy–Idaho Office (DOE-ID) and the INEEL operating contractor, Lockheed Martin Idaho Technologies Company (LMITCO). This incentive is worth $5 million dollars from FY-98 through FY-02 if the waste reduction goals are met. In addition, a second plan was prepared to show a path forward to either totally eliminate all radioactive liquid waste generation at INTEC by 2005 or find alternative waste treatment paths. Historically, this waste has been sent to an evaporator system with the bottoms sent to the INTEC Tank Farm. However, this Tank Farm is not RCRA permitted for mixed wastes and a Notice of Non-compliance Consent Order gives dates of 2003 and 2012 for removal of this waste from these tanks. Therefore, alternative treatments are needed for the waste streams. This plan investigated waste elimination opportunities as well as treatment alternatives. The alternatives, and the criteria for ranking these alternatives, were identified through Value Engineering meetings with all of the waste generators. The most promising alternatives were compared by applying weighting factors to each based on how well the alternative met the established criteria. From this information, an overall ranking of the various alternatives was obtained and a path forward recommended.

Tripp, Julia Lynn; Archibald, Kip Ernest; Argyle, Mark Don; Demmer, Ricky Lynn; Miller, Rose Anna; Lauerhass, Lance

1999-03-01T23:59:59.000Z

46

INEEL Radioactive Liquid Waste Reduction Program  

SciTech Connect

Reduction of radioactive liquid waste, much of which is Resource Conservation and Recovery Act (RCRA) listed, is a high priority at the Idaho National Technology and Engineering Center (INTEC). Major strides in the past five years have lead to significant decreases in generation and subsequent reduction in the overall cost of treatment of these wastes. In 1992, the INTEC, which is part of the Idaho National Environmental and Engineering Laboratory (INEEL), began a program to reduce the generation of radioactive liquid waste (both hazardous and non-hazardous). As part of this program, a Waste Minimization Plan was developed that detailed the various contributing waste streams, and identified methods to eliminate or reduce these waste streams. Reduction goals, which will reduce expected waste generation by 43%, were set for five years as part of this plan. The approval of the plan led to a Waste Minimization Incentive being put in place between the Department of Energy ? Idaho Office (DOE-ID) and the INEEL operating contractor, Lockheed Martin Idaho Technologies Company (LMITCO). This incentive is worth $5 million dollars from FY-98 through FY-02 if the waste reduction goals are met. In addition, a second plan was prepared to show a path forward to either totally eliminate all radioactive liquid waste generation at INTEC by 2005 or find alternative waste treatment paths. Historically, this waste has been sent to an evaporator system with the bottoms sent to the INTEC Tank Farm. However, this Tank Farm is not RCRA permitted for mixed wastes and a Notice of Non-compliance Consent Order gives dates of 2003 and 2012 for removal of this waste from these tanks. Therefore, alternative treatments are needed for the waste streams. This plan investigated waste elimination opportunities as well as treatment alternatives. The alternatives, and the criteria for ranking these alternatives, were identified through Value Engineering meetings with all of the waste generators. The most promising alternatives were compared by applying weighting factors to each based on how well the alternative met the established criteria. From this information, an overall ranking of the various alternatives was obtained and a path forward recommended.

C. B. Millet; J. L. Tripp; K. E. Archibald; L. Lauerhauss; M. D. Argyle; R. L. Demmer

1999-02-01T23:59:59.000Z

47

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

SciTech Connect

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.

COZZI, ALEX

2004-02-18T23:59:59.000Z

48

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

SciTech Connect

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.

COZZI, ALEX

2004-02-18T23:59:59.000Z

49

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

DOE Patents (OSTI)

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.

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

50

Process for removing sulfate anions from waste water  

DOE Patents (OSTI)

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.

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

51

Pilot studies to achieve waste minimization and enhance radioactive liquid waste treatment at the Los Alamos National Laboratory Radioactive Liquid Waste Treatment Facility  

SciTech Connect

The Radioactive and Industrial Wastewater Science Group manages and operates the Radioactive Liquid Waste Treatment Facility (RLWTF) at the Los Alamos National Laboratory (LANL). The RLWTF treats low-level radioactive liquid waste generated by research and analytical facilities at approximately 35 technical areas throughout the 43-square-mile site. The RLWTF treats an average of 5.8 million gallons (21.8-million liters) of liquid waste annually. Clarifloculation and filtration is the primary treatment technology used by the RLWTF. This technology has been used since the RLWTF became operable in 1963. Last year the RLWTF achieved an average of 99.7% removal of gross alpha activity in the waste stream. The treatment process requires the addition of chemicals for the flocculation and subsequent precipitation of radionuclides. The resultant sludge generated during this process is solidified in drums and stored or disposed of at LANL.

Freer, J.; Freer, E.; Bond, A. [and others

1996-07-01T23:59:59.000Z

52

The Savannah River Site's liquid radioactive waste operations involves the man  

NLE Websites -- All DOE Office Websites (Extended Search)

Site's liquid radioactive waste operations involves the management of space in the Site's Site's liquid radioactive waste operations involves the management of space in the Site's 49 underground waste tanks, including the removal of waste materials. Once water is removed from the waste tanks, two materials remain: salt and sludge waste. Removing salt waste, which fills approximately 90 percent of the tank space in the SRS tank farms, is a major step toward closing the Site's waste tanks that currently contain approximately 38 million gallons of waste. Due to the limited amount of tank space available in new-style tanks, some salt waste must be dispositioned in the interim to ensure sufficient tank space for continued sludge washing and to support the initial start-up and salt processing operations at the Salt Waste Processing Facility (SWPF).

53

Heterogeneous waste processing  

DOE Patents (OSTI)

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.

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

54

Tank Waste and Waste Processing | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Tank Waste and Waste Processing Tank Waste and Waste Processing Tank Waste and Waste Processing Tank Waste and Waste Processing The Defense Waste Processing Facility set a record by producing 267 canisters filled with glassified waste in a year. New bubbler technology and other enhancements will increase canister production in the future. The Defense Waste Processing Facility set a record by producing 267 canisters filled with glassified waste in a year. New bubbler technology and other enhancements will increase canister production in the future. A Savannah River Remediation employee uses a manipulator located inside a shielded enclosure at the Defense Waste Processing Facility where the melter is pouring molten glass inside a canister. A Savannah River Remediation employee uses a manipulator located inside a

55

Waste Management Process Improvement Project  

SciTech Connect

The Bechtel Hanford-led Environmental Restoration Contractor team's Waste Management Process Improvement Project is working diligently with the U.S. Department of Energy's (DOE) Richland Operations Office to improve the waste management process to meet DOE's need for an efficient, cost-effective program for the management of dangerous, low-level and mixed-low-level waste. Additionally the program must meet all applicable regulatory requirements. The need for improvement was highlighted when a change in the Groundwater/Vadose Zone Integration Project's waste management practices resulted in a larger amount of waste being generated than the waste management organization had been set up to handle.

Atwood, J.; Borden, G.; Rangel, G. R.

2002-02-25T23:59:59.000Z

56

November 8, 1983: Defense Waste Processing Facility | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

November 8, 1983: Defense Waste Processing Facility November 8, 1983: Defense Waste Processing Facility November 8, 1983: Defense Waste Processing Facility November 8, 1983: Defense Waste Processing Facility November 8, 1983 The Department begins construction of the Defense Waste Processing Facility (DWPF) at the Savannah River Plant in South Carolina. DWPF is designed to make high-level nuclear waste into a glass-like substance, which will then be shipped to a repository. DWPF will mix borosilicate glass with the waste, heat it to 2000 degrees F, and pour the mixture into stainless steel canisters. The mixture will cool into solid glass that can be permanently stored. DWPF will immobilize the more than 34 million gallons of liquid high-level waste that have accumulated from producing defense-related nuclear materials

57

Method for processing aqueous wastes  

DOE Patents (OSTI)

This invention is comprised of 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.

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

1992-12-31T23:59:59.000Z

58

Method for processing aqueous wastes  

DOE Patents (OSTI)

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.

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

1993-12-28T23:59:59.000Z

59

Method for processing aqueous wastes  

DOE Patents (OSTI)

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.

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

60

Oak Ridge National Laboratory TRU Waste Processing Center Tank Waste Processing Supernate Processing System  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

TRU Waste Processing Center TRU Waste Processing Center ORNL TRU Waste Processing Center Tank Waste Processing Supernate (SN) Processing System Presented by Don F. Gagel Vice President and Chief Technology Officer EnergX LLC ORNL TRU Waste Processing Center 1/21/09 2 SRS Technology Transfer, ORNL SN Process Overview SN Process Facility ORNL TRU Waste Processing Center 3 Waste Concentration Using Evaporator Evaporator Concentrates Waste Vapor stream superheated and HEPA-filtered Vapor stream exhausted to main ventilation system Supernate Pump and Evaporator Discharge Pump circulate waste between selected tank and evaporator during concentration. Evaporator Discharge Pump Supernate Pump Supernate Tank Evaporator Exhaust Blower ORNL TRU Waste Processing Center 4 Tank Sampling/ Transfer To Dryer Tank

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Characterization of industrial process waste heat and input heat streams  

SciTech Connect

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)

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

62

Salt Waste Processing Facility Fact Sheet  

Energy.gov (U.S. Department of Energy (DOE))

Nuclear material production operations at SRS resulted in the generation of liquid radioactive waste that is being stored, on an interim basis, in 49 underground waste storage tanks in the F- and H-Area Tank Farms.

63

Safety assessment of the liquid-fed ceramic melter process  

Science Conference Proceedings (OSTI)

As part of its development program for the solidification of high-level nuclear waste, Pacific Northwest Laboratory assessed the safety issues for a complete liquid-fed ceramic melter (LFCM) process. The LFCM process, an adaption of commercial glass-making technology, is being developed to convert high-level liquid waste from the nuclear fuel cycle into glass. This safety assessment uncovered no unresolved or significant safety problems with the LFCM process. Although in this assessment the LFCM process was not directly compared with other solidification processes, the safety hazards of the LFCM process are comparable to those of other processes. The high processing temperatures of the glass in the LFCM pose no additional significant safety concerns, and the dispersible inventory of dried waste (calcine) is small. This safety assessment was based on the nuclear power waste flowsheet, since power waste is more radioactive than defense waste at the time of solidification, and all accident conditions for the power waste would have greater radiological consequences than those for defense waste. An exhaustive list of possible off-standard conditions and equipment failures was compiled. These accidents were then classified according to severity of consequence and type of accident. Radionuclide releases to the stack were calculated for each group of accidents using conservative assumptions regarding the retention and decontamination features of the process and facility. Two recommendations that should be considered by process designers are given in the safety assessment.

Buelt, J.L.; Partain, W.L.

1980-08-01T23:59:59.000Z

64

EVALUATION OF ULTIMATE DISPOSAL METHOD FOR LIQUID AND SOLID RADIOACTIVE WASTES. PART I. INTERIM LIQUID STORAGE  

SciTech Connect

As the first part of a study to evaluate the economics of the various steps leading to and including the permanent disposal of high-activity liquid and solid radioactive waste, costs of interim liquid storage of acid and alkaline Purex and Thorex wastes were estimated for storage times of 0.5 to 30 years. A 6- ton/day plant was assumed, processing 1500 tons/year of uranium converter fuel at a burnup of 10,000 Mwd/ton and 270 tons/year of thorium converter fuel at a burnup of 20,000 Mwd/ton. Tanks of Savannah River design were assumed, with stainless steel construction for acid wastes and mild steel construction for neutralized wastes. The operating cycle of each tank was assumed to consist of equal filling and emptying periods plus a full (or dead) period. With interim storage time defined as filling time plus full time, tank costs were minimum when full time was 40 to 70% of the interim storage time, using present worth considerations. For waste storage times of 0.5 to 30 years, costs ranged from 2.2 x 10/sup -3/ to 9.5 x 10/sup -3/ mill/kwh/sub e/ for acid wastes and from 1.7 x 10/sup -3/ to 5.1 x 10/sup -3/ mill/kwh/sub e/ for neutralized wastes. (auth)

Bradshaw, R.L.; Perona, J.J.; Roberts, J.T.; Blomeke, J.O.

1961-08-22T23:59:59.000Z

65

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

SciTech Connect

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.

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

66

ICPP radioactive liquid and calcine waste technologies evaluation final report and recommendation  

SciTech Connect

Using a formalized Systems Engineering approach, the Latched Idaho Technologies Company developed and evaluated numerous alternatives for treating, immobilizing, and disposing of radioactive liquid and calcine wastes at the Idaho Chemical Processing Plant. Based on technical analysis data as of March, 1995, it is recommended that the Department of Energy consider a phased processing approach -- utilizing Radionuclide Partitioning for radioactive liquid and calcine waste treatment, FUETAP Grout for low-activity waste immobilization, and Glass (Vitrification) for high-activity waste immobilization -- as the preferred treatment and immobilization alternative.

1995-04-01T23:59:59.000Z

67

Future radioactive liquid waste streams study  

SciTech Connect

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.

Rey, A.S.

1993-11-01T23:59:59.000Z

68

Co-processing of agricultural and biomass waste with coal  

Science Conference Proceedings (OSTI)

A major thrust of our research program is the use of waste materials as co-liquefaction agents for the first-stage conversion of coal to liquid fuels. By fulfilling one or more of the roles of an expensive solvent in the direct coal liquefaction (DCL) process, the waste material is disposed off ex-landfill, and may improve the overall economics of DCL. Work in our group has concentrated on co-liquefaction with waste rubber tires, some results from which are presented elsewhere in these Preprints. In this paper, we report on preliminary results with agricultural and biomass-type waste as co-liquefaction agents.

Stiller, A.H.; Dadyburjor, D.B.; Wann, Ji-Perng [West Virginia Univ., Morgantown, WV (United States)] [and others

1995-12-31T23:59:59.000Z

69

Addressing mixed waste in plutonium processing  

SciTech Connect

The overall goal is the minimization of all waste generated in actinide processing facilities. Current emphasis is directed toward reducing and managing mixed waste in plutonium processing facilities. More specifically, the focus is on prioritizing plutonium processing technologies for development that will address major problems in mixed waste management. A five step methodological approach to identify, analyze, solve, and initiate corrective action for mixed waste problems in plutonium processing facilities has been developed.

Christensen, D.C.; Sohn, C.L. (Los Alamos National Lab., NM (United States)); Reid, R.A. (New Mexico Univ., Albuquerque, NM (United States). Anderson Schools of Management)

1991-01-01T23:59:59.000Z

70

Waste Heat Recovery from Industrial Process Heating Equipment -  

NLE Websites -- All DOE Office Websites (Extended Search)

Waste Heat Recovery from Industrial Process Heating Equipment - Waste Heat Recovery from Industrial Process Heating Equipment - Cross-cutting Research and Development Priorities Speaker(s): Sachin Nimbalkar Date: January 17, 2013 - 11:00am Location: 90-2063 Seminar Host/Point of Contact: Aimee McKane Waste heat is generated from several industrial systems used in manufacturing. The waste heat sources are distributed throughout a plant. The largest source for most industries is exhaust / flue gases or heated air from heating systems. This includes the high temperature gases from burners in process heating, lower temperature gases from heat treat, dryers, and heaters, heat from heat exchangers, cooling liquids and gases etc. The previous studies and direct contact with the industry as well as equipment suppliers have shown that a large amount of waste heat is not

71

Process development for remote-handled mixed-waste treatment  

SciTech Connect

The Oak Ridge National Laboratory (ORNL) is developing a treatment process for remote-handled (RH) liquid transuranic mixed waste governed by the concept of minimizing the volume of waste requiring disposal. This task is to be accomplished by decontaminating the bulk components so the process effluent can be disposed with less risk and expense. Practical processes have been demonstrated on the laboratory scale for removing cesium 137 and strontium 90 isotopes from the waste, generating a concentrated waste volume, and rendering the bulk of the waste nearly radiation free for downstream processing. The process is projected to give decontamination factors of 10{sup 4} for cesium and 10{sup 3} for strontium. Because of the extent of decontamination, downstream processing will be contact handled. The transuranic, radioactive fraction of the mixed waste stream will be solidified using a thin-film evaporator and/or microwave solidification system. Resultant solidified waste will be disposed at the Waste Isolation Pilot Plant (WIPP). 8 refs., 2 figs., 3 tabs.

Berry, J.B.; Campbell, D.O.; Lee, D.D.; White, T.L.

1990-01-01T23:59:59.000Z

72

System for recovering methane gas from liquid waste  

SciTech Connect

A system for and method of recovering methane gas from liquid waste which is stored within a pit is disclosed herein. The methane gas is produced by causing the liquid waste to undergo anaerobic fermentation. Therefore, it is necessary to close the pit in an air tight fashion. This is carried out using a cover sheet which is fixedly disposed over the pit in an air tight but readily disengagable fashion. The liquid waste within this air tight pit is preferably agitated intermittently during its storage therein whereby to increase the amount of methane gas produced.

Grabis, D.W.

1983-07-19T23:59:59.000Z

73

Detection of free liquid in containers of solidified radioactive waste  

DOE Patents (OSTI)

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.

Greenhalgh, Wilbur O. (Richland, WA)

1985-01-01T23:59:59.000Z

74

Polymer Solidification and Stabilization: Adaptable Processes for Atypical Wastes  

Science Conference Proceedings (OSTI)

Vinyl Ester Styrene (VES) and Advanced Polymer Solidification (APS{sup TM}) processes are used to solidify, stabilize, and immobilize radioactive, pyrophoric and hazardous wastes at US Department of Energy (DOE) and Department of Defense (DOD) sites, and commercial nuclear facilities. A wide range of projects have been accomplished, including in situ immobilization of ion exchange resin and carbon filter media in decommissioned submarines; underwater solidification of zirconium and hafnium machining swarf; solidification of uranium chips; impregnation of depth filters; immobilization of mercury, lead and other hazardous wastes (including paint chips and blasting media); and in situ solidification of submerged demineralizers. Discussion of the adaptability of the VES and APS{sup TM} processes is timely, given the decommissioning work at government sites, and efforts by commercial nuclear plants to reduce inventories of one-of-a-kind wastes. The VES and APS{sup TM} media and processes are highly adaptable to a wide range of waste forms, including liquids, slurries, bead and granular media; as well as metal fines, particles and larger pieces. With the ability to solidify/stabilize liquid wastes using high-speed mixing; wet sludges and solids by low-speed mixing; or bead and granular materials through in situ processing, these polymer will produce a stable, rock-hard product that has the ability to sequester many hazardous waste components and create Class B and C stabilized waste forms for disposal. Technical assessment and approval of these solidification processes and final waste forms have been greatly simplified by exhaustive waste form testing, as well as multiple NRC and CRCPD waste form approvals. (authors)

Jensen, C. [Diversified Technologies Services, Inc., Knoxville, TN (United States)

2007-07-01T23:59:59.000Z

75

Process for remediation of plastic waste  

DOE Patents (OSTI)

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.

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

2012-04-10T23:59:59.000Z

76

Process for remediation of plastic waste  

SciTech Connect

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.

Pol, Vilas G; Thiyagarajan, Pappannan

2013-11-12T23:59:59.000Z

77

Vibratory Shear Enhanced Process Filtration for Processing Decommissioning Wastes at Rancho Seco  

Science Conference Proceedings (OSTI)

Many non-nuclear industries use a vibratory shear enhanced filtration process (VSEP) to separate solids in liquid streams. Unlike other methods, including the application of a precoat of filter media, the VSEP does not generate any secondary waste, making it seem ideally suited for nuclear power plant radwaste systems. This report presents the results of laboratory and pilot scale in-plant testing of VSEP's ability to successfully process radioactive decommissioning waste. Testing at Rancho Seco showed t...

2003-12-02T23:59:59.000Z

78

Radioactive Liquid Waste Treatment Facility: Environmental Information Document  

Science Conference Proceedings (OSTI)

At Los Alamos National Laboratory (LANL), the treatment of radioactive liquid waste is an integral function of the LANL mission: to assure U.S. military deterrence capability through nuclear weapons technology. As part of this mission, LANL conducts nuclear materials research and development (R&D) activities. These activities generate radioactive liquid waste that must be handled in a manner to ensure protection of workers, the public, and the environment. Radioactive liquid waste currently generated at LANL is treated at the Radioactive Liquid Waste Treatment Facility (RLWTF), located at Technical Area (TA)-50. The RLWTF is 30 years old and nearing the end of its useful design life. The facility was designed at a time when environmental requirements, as well as more effective treatment technologies, were not inherent in engineering design criteria. The evolution of engineering design criteria has resulted in the older technology becoming less effective in treating radioactive liquid wastestreams in accordance with current National Pollutant Discharge Elimination System (NPDES) and Department of Energy (DOE) regulatory requirements. Therefore, to support ongoing R&D programs pertinent to its mission, LANL is in need of capabilities to efficiently treat radioactive liquid waste onsite or to transport the waste off site for treatment and/or disposal. The purpose of the EID is to provide the technical baseline information for subsequent preparation of an Environmental Impact Statement (EIS) for the RLWTF. This EID addresses the proposed action and alternatives for meeting the purpose and need for agency action.

Haagenstad, H.T.; Gonzales, G.; Suazo, I.L. [Los Alamos National Lab., NM (United States)

1993-11-01T23:59:59.000Z

79

The TEES process cleans waste and produces energy  

DOE Green Energy (OSTI)

A gasification system is under development that can be used with most types of wet organic wastes. The system operates at 350{degrees}C and 205 atm using a liquid water phase as the processing medium. Since a pressurized system is used, the wet waste can be fed as a solution or slurry to the reactor without drying. Through the development of catalysts, a useful processing system has been produced. The system has utility both for direct conversion of high-moisture biomass to fuel gas or as a wastewater cleanup system for wet organic wastes including unconverted biomass from bioconversion processes. By the use of this system >99% conversions of organic waste to medium-Btu fuel gas can be achieved.

Elliott, D.C.; Silva, L.J.

1995-02-01T23:59:59.000Z

80

Process for treating fission waste  

DOE Patents (OSTI)

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.

Rohrmann, Charles A. (Kennewick, WA); Wick, Oswald J. (Richland, WA)

1983-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Office of Waste Processing Technical Exchange  

NLE Websites -- All DOE Office Websites (Extended Search)

Contact Information: For more EM Waste Processing Technical Exchange 2010 information, please contact one of the folowing (click name to email): Bill Wilmarth Rosalind Blocker...

82

Office of Waste Processing Technical Exchange  

... Savannah River/Hanford/Idaho along with others receiving funding from the Environmental Management Office of Waste Processing have met to exchange ...

83

Office of Waste Processing Technical Exchange  

... Savannah River/Hanford/Idaho along with others receiving funding from the Environmental Management Office of Waste Processing have met to exchange recent ...

84

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

DOE Green Energy (OSTI)

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.

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

2008-12-01T23:59:59.000Z

85

HNPF LIQUID WASTE DISPOSAL COST STUDY  

SciTech Connect

The HNPF cost analysis for waste disposal was made on the basis of 10,000 gallons of laundry waste and 9,000 gallons of other plant waste per year. The costs are compared for storage at HNPF site for 10 yr, packaging and shipment to AEC barial ground, packaging and shipment for sea disposal, and disposal by licensed vendor. A graphical comparison is given for the yearly costs of disposal by licensed vendor and the evaporator system as a function of waste volume. Recommendations are included for the handling of the wastes expected from HNPF operations. (B.O.G.)

Piccot, A.R.

1959-11-01T23:59:59.000Z

86

Liquid and Gaseous Waste Operations Project Annual Operating Report CY 1999  

SciTech Connect

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.

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

2000-03-01T23:59:59.000Z

87

RECOVERY OF MERCURY FROM CONTAMINATED LIQUID WASTES  

SciTech Connect

Mercury was widely used in U.S. Department of Energy (DOE) weapons facilities, resulting in a broad range of mercury-contaminated wastes and wastewaters. Some of the mercury contamination has escaped to the local environment, particularly at the Y-12 Plant in Oak Ridge, Tennessee, where approximately 330 metric tons of mercury were discharged to the environment between 1953 and 1963 (TN & Associates, 1998). Effective removal of mercury contamination from water is a complex and difficult problem. In particular, mercury treatment of natural waters is difficult because of the low regulatory standards. For example, the Environmental Protection Agency has established a national ambient water quality standard of 12 parts-per-trillion (ppt), whereas the standard is 1.8 ppt in the Great Lakes Region. In addition, mercury in the environment is typically present in several different forms, but sorption processes are rarely effective with more than one or two of these forms. To meet the low regulatory discharge limits, an effective sorption process must be able to address all forms of mercury present in the water. One approach is to apply different sorbents in series depending on the mercury speciation and the regulatory discharge limits. ADA Technologies, Inc. has developed four new sorbents to address the variety of mercury species present in industrial discharges and natural waters. Three of these sorbents have been field tested on contaminated creek water at the Y-12 Plant. Two of these sorbents have been successfully demonstrated very high removal efficiencies for soluble mercury species, reducing mercury concentrations at the outlet of a pilot-scale system to less than 12 ppt for as long as six months. The other sorbent tested at the Y-12 Plant targeted colloidal mercury not removed by standard sorption or filtration processes. At the Y-12 Plant, colloidal mercury appears to be associated with iron, so a sorbent that removes mercury-iron complexes in the presence of a magnetic field was evaluated. Field results indicated good removal of this mercury fraction from the Y-12 waters. In addition, this sorbent is easily regenerated by simply removing the magnetic field and flushing the columns with water. The fourth sorbent is still undergoing laboratory development, but results to date indicate exceptionally high mercury sorption capacity. The sorbent is capable of removing all forms of mercury typically present in natural and industrial waters, including Hg{sup 2+}, elemental mercury, methyl mercury, and colloidal mercury. The process possesses very fast kinetics, which allows for higher flow rates and smaller treatment units. These sorbent technologies, used in tandem or individually depending on the treatment needs, can provide DOE sites with a cost-effective method for reducing mercury concentrations to very low levels mandated by the regulatory community. In addition, the technologies do not generate significant amounts of secondary wastes for disposal. Furthermore, the need for improved water treatment technologies is not unique to the DOE. The new, stringent requirements on mercury concentrations impact other government agencies as well as the private sector. Some of the private-sector industries needing improved methods for removing mercury from water include mining, chloralkali production, chemical processing, and medical waste treatment. The next logical step is to deploy one or more of these sorbents at a contaminated DOE site or at a commercial facility needing improved mercury treatment technologies. A full-scale deployment is planned in fiscal year 2000.

Robin M. Stewart

1999-09-29T23:59:59.000Z

88

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

SciTech Connect

A Generic Technology for treatment of DOE Metal-Bearing Liquid Waste The DOE metal-bearing liquid waste inventory is large and diverse, both with respect to the metals (heavy metals, transuranics, radionuclides) themselves, and the nature of the other species (annions, organics, etc.) present. Separation and concentration of metals is of interest from the standpoint of reducing the volume of waste that will require special treatment or isolation, as well as, potentially, from the standpoint of returning some materials to commerce by recycling. The variety of metal-bearing liquid waste in the DOE complex is so great that it is unlikely that any one process (or class of processes) will be suitable for all material. However, processes capable of dealing with a wide variety of wastes will have major advantages in terms of process development, capital, and operating costs, as well as in environmental and safety permitting. Moreover, to the extent that a process operates well with a variety of metal-bearing liquid feedwastes, its performance is likely to be relatively robust with respect to the inevitable composition variations in each waste feed. One such class of processes involves high-temperature treatment of atomized liquid waste to promote reactive capture of volatile metallic species on collectible particulate substrates injected downstream of a flame zone. Compared to low-temperature processes that remove metals from the original liquid phase by extraction, precipitation, ion exchange, etc., some of the attractive features of high-temperature reactive scavenging are: The organic constituents of some metal-bearing liquid wastes (in particular, some low-level mixed wastes) must be treated thermally in order to meet the requirements of the Resource Conservation and Recovery Act (RCRA) and Toxic Substances Control Act (TSCA), and the laws of various states. No species need be added to an already complex liquid system. This is especially important in light of the fact that DOE has already experienced problems with organic complexants added to precipitate radionuclides. For example, the Defense Nuclear Facilities Safety Board has expressed, in a formal Recommendation to the Secretary of Energy, its concern about the evolution of benzene vapor in concentrations greater then the lower flammability limit from tanks to which sodium tetraphenylborate has been added to precipitate 137Cs in the ''In-Tank Precipitation'' (ITP) process at the Savannah River Site. Other species added to the waste in the ITP process are sodium titanate (to adsorb 90Sr and Pu), and oxalic acid. Avoiding addition of organics to radioactive waste has the additional advantage that is likely to significantly reduce the rate of radiolytic and radiolytically-induced hydrogen generation (c.f. Meisel et al., [1993]), in which it is shown that removal of organics reduces the rate of hydrogen generation in simulated waste from Hanford tank 241-SY-101 by over 70%. Organic species already present are destroyed with very high efficiency. This attribute is especially attractive with respect to high-level tank waste at the Hanford Site, in which large amounts of citrate, glyoxylate, EDTA (ethylenediaminetetraacetic acid), and HEDTA [N-(2- hydroxyethyl)-ethylenediaminetriacetic acid] were added to precipitate radionuclides. These organic species are important in the thermal and radiolytic generation of methane, hydrogen, and nitrous oxide, flammable mixtures of which are episodically vented from 25 tanks on Hanford's Flammable Gas Watch List [Hopkins, 1994]. The same basic approach can be used to treat a broad range of liquid wastes, in each case concentrating the metals (regardless of liquid-phase oxidation state or association with chelators or absorbents) using a collectible sorbent, and destroying any organic species present. In common with the Army's approach (see section 2.2) to the thermal destruction of a 10 range of chemical warfare agents (GB, VX, and two blister agents), this may drastically simplify process and plant design and

Wendt, Jost O.L.

2001-09-30T23:59:59.000Z

89

ADSORPTION SEPARATION PROCESSES FOR IONIC LIQUID CATALYTIC ...  

Presently disclosed are methods and apparatus for separation of reaction products from reaction mixtures in an ionic liquid catalysis process, particularly in ...

90

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

SciTech Connect

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.

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

1998-06-18T23:59:59.000Z

91

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

SciTech Connect

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.

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

92

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

SciTech Connect

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.

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

93

Process for remediation of plastic waste - Energy Innovation ...  

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 ...

94

Waste Processing Annual Technology Development Report 2007  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Processing Processing Annual Technology Development Report 2007 SRNS-STI-2008-00040 United States Department of Energy Waste Processing Annual Technology Development Report 2007 Prepared and edited by S. R. Bush EM Technical Integration Office Savannah River National Laboratory Reviewed by Dr. W. R. Wilmarth, Manager EM Technical Integration Office Savannah River National Laboratory Approved by Dr. S. L. Krahn, Director EM-21 Office of Waste Processing U. S. Department of Energy APPROVED for Release for Unlimited (Release to Public) (Signed 08/13/2008) (Signed 08/13/2008) (Signed 08/13/2008) EM-21 Waste Processing Annual Report for Calendar Year 2007 2/74

95

Using Waste Heat for External Processes  

Science Conference Proceedings (OSTI)

This DOE Industrial Technologies Program tip sheet describes the savings resulting from using waste heat from high-temperature industrial processes for lower temperature processes, like oven-drying.

Not Available

2006-01-01T23:59:59.000Z

96

Recovery of valuable materials from waste liquid crystal display panel  

Science Conference Proceedings (OSTI)

Associated with the rapid development of the information and electronic industry, liquid crystal displays (LCDs) have been increasingly sold as displays. However, during the discarding at their end-of-life stage, significant environmental hazards, impacts on health and a loss of resources may occur, if the scraps are not managed in an appropriate way. In order to improve the efficiency of the recovery of valuable materials from waste LCDs panel in an environmentally sound manner, this study presents a combined recycling technology process on the basis of manual dismantling and chemical treatment of LCDs. Three key processes of this technology have been studied, including the separation of LCD polarizing film by thermal shock method the removal of liquid crystals between the glass substrates by the ultrasonic cleaning, and the recovery of indium metal from glass by dissolution. The results show that valuable materials (e.g. indium) and harmful substances (e.g. liquid crystals) could be efficiently recovered or separated through above-mentioned combined technology. The optimal conditions are: (1) the peak temperature of thermal shock to separate polarizing film, ranges from 230 to 240 deg. C, where pyrolysis could be avoided; (2) the ultrasonic-assisted cleaning was most efficient at a frequency of 40 KHz (P = 40 W) and the exposure of the substrate to industrial detergents for 10 min; and (3) indium separation from glass in a mix of concentrated hydrochloric acid at 38% and nitric acid at 69% (HCl:HNO{sub 3}:H{sub 2}O = 45:5:50, volume ratio). The indium separation process was conducted with an exposure time of 30 min at a constant temperature of 60 deg. C.

Li Jinhui [Department of Environmental Science and Engineering, Tsinghua University (China); Sino-Italia Environmental Energy Building, Room 804, Haidian District, Beijing 100084 (China)], E-mail: jinhui@tsinghua.edu.cn; Gao Song; Duan Huabo; Liu Lili [Department of Environmental Science and Engineering, Tsinghua University (China)

2009-07-15T23:59:59.000Z

97

Assessment of Tank 241-S-112 Liquid Waste Mixing in Tank 241-SY-101  

SciTech Connect

The objectives of this study were to evaluate mixing of liquid waste from Tank 241-S-112 with waste in Tank 241-SY-101 and to determine the properties of the resulting waste for the cross-site transfer to avoid potential double-shell tank corrosion and pipeline plugging. We applied the time-varying, three-dimensional computer code TEMPEST to Tank SY-101 as it received the S-112 liquid waste. The model predicts that temperature variations in Tank SY-101 generate a natural convection flow that is very slow, varying from about 7 x 10{sup -5} to 1 x 10{sup -3} ft/sec (0.3 to about 4 ft/hr) in most areas. Thus, natural convection would eventually mix the liquid waste in SY-101 but would be very slow to achieve nearly complete mixing. These simulations indicate that the mixing of S-112 and SY-101 wastes in Tank SY-101 is a very slow process, and the density difference between the two wastes would further limit mixing. It is expected to take days or weeks to achieve relatively complete mixing in Tank SY-101.

Onishi, Yasuo; Trent, Donald S.; Wells, Beric E.; Mahoney, Lenna A.

2003-10-01T23:59:59.000Z

98

Calcination process for radioactive wastes  

DOE Patents (OSTI)

The present invention provides a method for minimizing the volatilization of chlorides during solidification in a fluidized-bed calciner of liquids containing sodium, nitrate and chloride ions. Zirconium and fluoride are introduced into the liquid, and one-half mole of calcium nitrate is added per mole of fluoride present in the liquid mixture. The mixture is calcined in the fluidized-bed calciner at about 500.degree.C., producing a high bulk density calcine product containing the chloride, thus tying up the chloride in the solid product and minimizing chloride volatilization.

Kilian, Douglas C. (Kennewick, WA)

1976-05-04T23:59:59.000Z

99

Role of Liquid Waste Pretreatment Technologies in Solving the DOE Clean-up Mission  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Role of Liquid Waste Pretreatment Technologies in Role of Liquid Waste Pretreatment Technologies in Solving the DOE Clean-up Mission W. R. Wilmarth March 5 2009 March 5, 2009 HLW Corporate Board Phoenix AZ HLW Corporate Board, Phoenix, AZ Co-authors M. E. Johnson, CH2M Hill Plateau Remediation Company G. Lumetta, Pacific Northwest National Laboratory N Machara DOE Office of Engineering and Technology N. Machara, DOE Office of Engineering and Technology M. R. Poirier, Savannah River National Laboratory P C S DOE S h Ri P. C. Suggs, DOE Savannah River M. C. Thompson, Savannah River National Laboratory, Retired Retired 2 Background Separations is a fundamental business within DOE. The role of separations today is to expedite waste retrieval The role of separations today is to expedite waste retrieval, processing and closure. Recognized as part of E&T Roadmap

100

Method for co-processing waste rubber and carbonaceous material  

DOE Green Energy (OSTI)

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.

Farcasiu, M.; Smith, C.M.

1990-10-09T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Method for co-processing waste rubber and carbonaceous material  

DOE Green Energy (OSTI)

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.

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

1991-01-01T23:59:59.000Z

102

Waste Heat Management Options: Industrial Process Heating Systems  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Heat Management Options Heat Management Options Industrial Process Heating Systems By Dr. Arvind C. Thekdi E-mail: athekdi@e3minc.com E3M, Inc. August 20, 2009 2 Source of Waste Heat in Industries * Steam Generation * Fluid Heating * Calcining * Drying * Heat Treating * Metal Heating * Metal and Non-metal Melting * Smelting, agglomeration etc. * Curing and Forming * Other Heating Waste heat is everywhere! Arvind Thekdi, E3M Inc Arvind Thekdi, E3M Inc 3 Waste Heat Sources from Process Heating Equipment * Hot gases - combustion products - Temperature from 300 deg. F. to 3000 deg.F. * Radiation-Convection heat loss - From temperature source of 500 deg. F. to 2500 deg. F. * Sensible-latent heat in heated product - From temperature 400 deg. F. to 2200 deg. F. * Cooling water or other liquids - Temperature from 100 deg. F. to 180 deg. F.

103

Process and installation for simultaneously producing compost and biogas from organic waste  

Science Conference Proceedings (OSTI)

A process is described for the simultaneous treatment of solid or semi-solid organic waste and liquid organic waste with a view to the simultaneous production of compost and biogas, wherein the liquid organic waste is subjected to a liquid-solid separation. The liquid phase from this separation is subjected to anaerobic fermentation in at least one closed digester, the solid phase from the liquid-solid separation is mixed with the solid or semi-solid organic waste, and the resulting mixture is subjected to aerobic fermentation at the periphery of the digester and in contact therewith. Mud, clarified liquid and gas are respectively discharged from the digester whereas compost from the aerobic fermentation of the solid or semi-solid waste is recovered at the periphery of the digester wherein the digester is characterized by two superimposed compartments, an upper compartment at low pressure and a lower compartment at high pressure, the compartments communicating together through at least one lateral pipe and through a central siphon. A means is provided for lowering the pressure of the lower compartment when the liquid reaches a predetermined level therein. An installation is described for the simultaneous treatment of solid or semi-solid organic waste and liquid waste with a view to the simultaneous production of compost and biogas. This comprises: means for separating the liquid organic waste into a solid phase and a liquid phase; at least one closed digester; means for introducing the liquid phase into the digester; means for mixing the solid phase with the solid or semi-solid waste; means for bringing the resulting mixture to the periphery of the digester in contact therewith; and means for discharging respectively from the digester the gas which is formed therein by anaerobic fermentation and the sludges which are deposited therein.

Lebesgue, Y.; Zeana, A.

1986-12-30T23:59:59.000Z

104

Liquid waste certification plan 340 waste handling facility  

Science Conference Proceedings (OSTI)

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

HALGREN, D.L.

1999-04-21T23:59:59.000Z

105

Office of Waste Processing Technical Exchange  

NLE Websites -- All DOE Office Websites (Extended Search)

this hotel at the government per diem rate of 132.00 per night. Please reference the "DOE EM Waste Processing Technical Exchange 2010" when making your reservation to the get...

106

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Oak Ridge Transuranic Waste Processing Center, September 2013 September 2013 Review of Management of Safety Systems at the Oak Ridge Transuranic Waste Processing Center and...

107

Haze Formation and Behavior in Liquid-Liquid Extraction Processes  

Science Conference Proceedings (OSTI)

Aqueous haze formation and behavior was studied in the liquid-liquid system tri-n-butyl phosphate in odorless kerosene and 3M nitric acid with uranyl nitrate and cesium nitrate representing the major solute and an impurity, respectively. A pulsed column, mixer-settler and centrifugal contactor were chosen to investigate the effect of different turbulence characteristics on the manifestation of haze since these contactors exhibit distinct mixing phenomena. The dispersive processes of drop coalescence and breakage, and water precipitation in the organic phase were observed to lead to the formation of haze drops of {approx}1 um in diameter. The interaction between the haze and primary drops of the dispersion was critical to the separation efficiency of the liquid-liquid extraction equipment. Conditions of high power input and spatially homogeneous mixing enabled the haze drops to become rapidly assimilated within the dispersion to maximize the scrub performance and separation efficiency of the equipment.

Arm, Stuart T.; Jenkins, J. A.

2006-07-31T23:59:59.000Z

108

Iraq liquid radioactive waste tanks maintenance and monitoring program plan.  

SciTech Connect

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.

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

2011-10-01T23:59:59.000Z

109

Process for treating alkaline wastes for vitrification  

DOE Patents (OSTI)

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.

Hsu, Chia-lin W.

1994-01-01T23:59:59.000Z

110

Decontamination processes for waste glass canisters  

SciTech Connect

The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO/sub 3/-HF and H/sub 2/C/sub 2/O/sub 4/ to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated.

Rankin, W.N.

1981-06-01T23:59:59.000Z

111

Exploratory study of complexant concentrate waste processing  

SciTech Connect

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.

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

1993-02-01T23:59:59.000Z

112

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

SciTech Connect

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.

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

1999-10-01T23:59:59.000Z

113

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

SciTech Connect

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.

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

1999-09-30T23:59:59.000Z

114

Low-level liquid waste treatment system start-up  

Science Conference Proceedings (OSTI)

Following removal of Cs-137 by ion exchange in the Supernatant Treatment System immediately upstream, the radioactive liquid waste is volume-reduced by evaporation. Trace amounts of Cs-137 in the resulting distillate are removed by ion exchange, then the distillate is discharged to the existing plant water treatment system. The concentrated product, 37 to 41 percent solids (by weight), is encapsulated in cement, producing a stable low-level waste form. This report provides a summary of work performed to test the Liquid Waste Treatment System following construction turnover and prior to radioactive operation. All mechanical and electrical components, piping, valves, pumps, tanks, controls, and instrumentation required to operate the system were tested; first with water, then with simulated waste. Subsystems (individual tanks, pumps, and control loops) were tested individually, then as a complete system. Finally, the system began a controlled start-up phase, which included the first four months of radioactive operation. Components were tested for operability then for performance data to verify the system`s ability to produce an acceptable waste form at design feed rates.

Baker, M.N.; Gessner, R.F.

1989-07-01T23:59:59.000Z

115

Process for preparing a liquid fuel composition  

SciTech Connect

A process for preparing a liquid fuel composition which comprises liquefying coal, separating a mixture of phenols from said liquefied coal, converting said phenols to the corresponding mixture of anisoles, subjecting at least a portion of the remainder of said liquefied coal to hydrotreatment, subjecting at least a portion of said hydrotreated liquefied coal to reforming to obtain reformate and then combining at least a portion of said anisoles and at least a portion of said reformate to obtain said liquid fuel composition.

Singerman, Gary M. (Monroeville, PA)

1982-03-16T23:59:59.000Z

116

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

SciTech Connect

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.

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

117

Office of Waste Processing Technical Exchange  

NLE Websites -- All DOE Office Websites (Extended Search)

Event Media Links Event Media Links Session 1: Technical Exchange Opening Topic Speaker PDF Podcast S01-01 Welcome T. Michalske, SRNL N/A Podcast S01-03 Introductions G. Flowers, SRNS N/A Podcast S01-04 Opening Remarks I. Triay, DOE-EM Presentation PDF Podcast S01-05 Status of Waste Processing Technology Development S. Schneider, DOE-EM Presentation PDF Podcast S01-06 Hanford/SRS Tank Waste Path Forward K. Subramanian/ T. Sams, SRR/WRPS Presentation PDF Podcast S01-07 Fluidized Bed Steam Reformer Overview B. Mason, TTT Presentation PDF Podcast S01-08 Next Generation Cesium Solvent B.Moyer/S. Fink/M. Geeting, ORNL/SRNL/SRR Presentation PDF Podcast S01-09 Rotary Microfilter Development/Small Column Ion Exchange D. Herman/ R. Edwards, SRNL/SRR Presentation PDF Podcast Session 2: Increased Waste Loading - Improved Current Processing

118

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

SciTech Connect

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.

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

1985-01-01T23:59:59.000Z

119

Improved Consolidation Process for Producing Ceramic Waste forms  

DOE Patents (OSTI)

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.

Hash, Harry C.; Hash, Mark C.

1998-07-24T23:59:59.000Z

120

Office of Waste Processing Technical Exchange  

NLE Websites -- All DOE Office Websites (Extended Search)

EM Waste Processing Technical Exchange 2010 Agenda EM Waste Processing Technical Exchange 2010 Agenda (Sponsored by EM Office of Waste Processing) November 16 - 18, 2010; Loews Hotel, Atlanta, GA 11/2/2010 Monday, November 15, 2010 5:00 - 7:00 pm Early Registration and Speaker Check-in *Light Refreshments Tuesday Morning, November 16, 2010 Session 1: Technical Exchange Opening (Chair: W. Wilmarth); Salon D Live Webcast Click the video icon to view Session 1 Live Webcast Submit Question Click the Question icon to submit a question. Time Topic Speaker 7:00 am Registration and Check-in 8:00 am S01-01 Welcome T. Michalske, SRNL 8:05 am S01-02 Opening Comments Y. Collazo, DOE-EM 8:15 am S01-03 Introductions G. Flowers, SRNS 8:20 am S01-04 Opening Remarks I. Triay, DOE-EM 8:45 am S01-05 Status of Waste Processing Technology Development

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Microsoft PowerPoint - S05-07_Varona_Solid-Liquid Waste Interface.pptx  

NLE Websites -- All DOE Office Websites (Extended Search)

Liquid Interface Monitor Liquid Interface Monitor (SLIM) Jose Varona D. Roelant, A. Awwad, D. McDaniel Florida International University's Applied Research Center EM Waste Processing Technical Exchange November 17, 2010 Print Close Disclaimer This presentation was prepared as an account of work sponsored by an agency of the United States government (Department of Energy, Office of Environmental Management, under Grant No. DE-FG01-05EW07033). Neither the United States government nor any agency thereof, nor any of their employees, nor any of its contractors, subcontractors, nor their employees makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness,

122

Improved FGD dewatering process cuts solid wastes  

Science Conference Proceedings (OSTI)

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.

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

2009-08-15T23:59:59.000Z

123

New Facility Saves $20 Million, Accelerates Waste Processing | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Facility Saves $20 Million, Accelerates Waste Processing Facility Saves $20 Million, Accelerates Waste Processing New Facility Saves $20 Million, Accelerates Waste Processing August 15, 2012 - 12:00pm Addthis The new Cask Processing Enclosure (CPE) facility is located at the Transuranic Waste Processing Center (TWPC). The Transuranic Waste Processing Center (TWPC) processes, repackages, and ships the site's legacy TRU waste offsite. OAK RIDGE, Tenn. - Oak Ridge's EM program recently began operations at a newly constructed facility that will accelerate the completion of remote-handled transuranic (TRU) waste processing at the site by two years and save taxpayers more than $20 million. The new Cask Processing Enclosure (CPE) facility is located at the Transuranic Waste Processing Center (TWPC). TWPC processes, repackages, and

124

Process for treating alkaline wastes for vitrification  

DOE Patents (OSTI)

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.

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

1995-01-01T23:59:59.000Z

125

Process and system for treating waste water  

DOE Patents (OSTI)

A process of treating raw or primary waste water using a powdered, activated carbon/aerated biological treatment system is disclosed. Effluent turbidities less than 2 JTU (Jackson turbidity units), zero TOC (total organic carbon) and in the range of 10 mg/l COD (chemical oxygen demand) can be obtained. An influent stream of raw or primary waste water is contacted with an acidified, powdered, activated carbon/alum mixture. Lime is then added to the slurry to raise the pH to about 7.0. A polyelectrolyte flocculant is added to the slurry followed by a flocculation period -- then sedimentation and filtration. The separated solids (sludge) are aerated in a stabilization sludge basin and a portion thereof recycled to an aerated contact basin for mixing with the influent waste water stream prior to or after contact of the influent stream with the powdered, activated carbon/alum mixture.

Olesen, Douglas E. (Kennewick, WA); Shuckrow, Alan J. (Pasco, WA)

1978-01-01T23:59:59.000Z

126

Techniques and Facilities for Handling and Packaging Tritiated Liquid Wastes for Burial  

SciTech Connect

Methods and facilities have been developed for the collection, storage, measurement, assay, solidification, and packaging of tritiated liquid wastes (concentrations up to 5 Ci/ml) for disposal by land burial. Tritium losses to the environment from these operations are less than 1 ppm. All operations are performed in an inert gas-purged glovebox system vented to an effluent removal system which permits nearly complete removal of tritium from the exhaust gases prior to their dischardge to the environment. Waste oil and water from tritium processing areas are vacuum-transferred to glovebox storage tanks through double-walled lines. Accommodations are also available for emptying portable liquid waste containers and for removing tritiated water from molecular sieve beds with heat and vacuum. The tritium concentration of the collected liquids is measured by an in-line calorimeter. A low-volume metering pump is used to transfer liquids from holding tanks to heavy walled polyethylene drums filled with an absorbent or cement for solidification. Final packaging of the sealed polyethylene drums is in either an asphalt-filled combination 30- and 55- gallon metal drum package or a 30-gallon welded stainless steel container.

Rhinehammer, T. B.; Mershad, E. A.

1974-06-01T23:59:59.000Z

127

Separating and Stabilizing Phosphate from High-Level Radioactive Waste: Process Development and Spectroscopic Monitoring  

SciTech Connect

Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

Lumetta, Gregg J.; Braley, Jenifer C.; Peterson, James M.; Bryan, Samuel A.; Levitskaia, Tatiana G.

2012-05-09T23:59:59.000Z

128

Metal decontamination for waste minimization using liquid metal refining technology  

Science Conference Proceedings (OSTI)

The current Department of Energy Mixed Waste Treatment Project flowsheet indicates that no conventional technology, other than surface decontamination, exists for metal processing. Current Department of Energy guidelines require retrievable storage of all metallic wastes containing transuranic elements above a certain concentration. This project is in support of the National Mixed Low Level Waste Treatment Program. Because of the high cost of disposal, it is important to develop an effective decontamination and volume reduction method for low-level contaminated metals. It is important to be able to decontaminate complex shapes where surfaces are hidden or inaccessible to surface decontamination processes and destruction of organic contamination. These goals can be achieved by adapting commercial metal refining processes to handle radioactive and organic contaminated metal. The radioactive components are concentrated in the slag, which is subsequently vitrified; hazardous organics are destroyed by the intense heat of the bath. The metal, after having been melted and purified, could be recycled for use within the DOE complex. In this project, we evaluated current state-of-the-art technologies for metal refining, with special reference to the removal of radioactive contaminants and the destruction of hazardous organics. This evaluation was based on literature reports, industrial experience, plant visits, thermodynamic calculations, and engineering aspects of the various processes. The key issues addressed included radioactive partitioning between the metal and slag phases, minimization of secondary wastes, operability of the process subject to widely varying feed chemistry, and the ability to seal the candidate process to prevent the release of hazardous species.

Joyce, E.L. Jr.; Lally, B. [Los Alamos National Lab., NM (United States); Ozturk, B.; Fruehan, R.J. [Carnegie-Mellon Univ., Pittsburgh, PA (United States). Dept. of Materials Science and Engineering

1993-09-01T23:59:59.000Z

129

Idaho Chemical Processing Plant spent fuel and waste management technology development program plan: 1994 Update  

SciTech Connect

The Department of Energy has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage since 1951 and reprocessing since 1953. Until April 1992, the major activity of the ICPP was the reprocessing of SNF to recover fissile uranium and the management of the resulting high-level wastes (HLW). In 1992, DOE chose to discontinue reprocessing SNF for uranium recovery and shifted its focus toward the continued safe management and disposition of SNF and 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), 3,800 cubic meters of calcine waste, and 289 metric tons heavy metal of SNF are in inventory at the ICPP. Disposal of SNF and high-level waste (HLW) is planned for a repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP spent Fuel and Waste Management Technology Development Program (SF&WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will be properly stored and prepared for final disposal in accordance with regulatory drivers. This Plan presents a brief summary of each of the major elements of the SF&WMTDP; identifies key program assumptions and their bases; and outlines the key activities and decisions that must be completed to identify, develop, demonstrate, and implement a process(es) that will properly prepare the SNF and radioactive wastes stored at the ICPP for safe and efficient interim storage and final disposal.

1994-09-01T23:59:59.000Z

130

Waste Form Features, Events, and Processes  

Science Conference Proceedings (OSTI)

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

R. Schreiner

2004-10-27T23:59:59.000Z

131

Office of Waste Processing Technical Exchange  

NLE Websites -- All DOE Office Websites (Extended Search)

Agenda Hotel Register Contacts Event Media Speaker Information Home Agenda Hotel Register Contacts Event Media Speaker Information Home Environmental Management Waste Processing Technical Exchange 2010 in Atlanta, GA, November 16 - 18. Over the past eight years, personnel from the three sites, Savannah River/Hanford/Idaho along with others receiving funding from the Environmental Management Office of Waste Processing have met to exchange recent results of on-going field operations and technology development. The purpose of this exchange is to provide a forum for discussion of each Site's efforts to accelerate cleanup operations. Keys to success and lessons learned are openly exchanged in a manner to allow for open discussion between operations, engineering and scientists to accelerate transition of technologies from concepts to field implementation.

132

Enzymes and microorganisms in food industry waste processing and conversion to useful products: a review of the literature  

DOE Green Energy (OSTI)

Bioconversion of food processing wastes is receiving increased attention with the realization that waste components represent an available and utilizable resource for conversion to useful products. Liquid wastes are characterized as dilute streams containing sugars, starches, proteins, and fats. Solid wastes are generally cellulosic, but may contain other biopolymers. The greatest potential for economic bioconversion is represented by processes to convert cellulose to glucose, glucose to alcohol and protein, starch to invert sugar, and dilute waste streams to methane by anaerobic digestion. Microbial or enzymatic processes to accomplish these conversions are described.

Carroad, P.A.; Wilke, C.R.

1976-12-01T23:59:59.000Z

133

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

Energy.gov (U.S. Department of Energy (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...

134

Liquid Metal Processing and Casting 2013  

Science Conference Proceedings (OSTI)

Ceramic, Slag and Refractory Reactions with Liquid Metals - Refining, Evaporation and Gas/Metal Reactions - Fundamentals of Reactions involving Liquid ...

135

Modified Bayer Process for Alumina Removal from Hanford Waste  

AREVA NC Inc. Modified Bayer Process for Alumina Removal from Hanford Waste January 24, 2007 Don Geniesse AREVA NC Inc.

136

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

SciTech Connect

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.

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

137

LFCM (liquid-fed ceramic melter) processing characteristics of mercury  

SciTech Connect

An experimental-scale liquid-fed ceramic melter was used in a series of tests to evaluate the processing characteristics of mercury in simulated defense waste under various melter operating conditions. This solidification technology had no detectable capacity for incorporating mercury into its borosilicate, vitreous, product, and essentially all the mercury fed to the melter was lost to the off-gas system as gaseous effluent. An ejector venturi scrubber condensed and collected 97% of the mercury evolved from the melter. Chemically the condensed mercury effluent was composed entirely of chlorides, and except in a low-temperature test, mercury chlorides (Hg{sub 2}Cl{sub 2}) was the primary chloride formed. As a result, combined mercury accounted for most of the insoluble mass collected by the process quench scrubber. Although macroscopic quantities of elemental mercury were never observed in process secondary waste streams, finely divided and dispersed mercury that blackened all condensed Hg{sub 2}Cl{sub 2} residues was capable of saturating the quenched process exhaust with mercury vapor. However, the vapor pressure of mercury in the quenched melter exhaust was easily and predictably controlled with an off-gas stream chiller. 5 refs., 4 figs., 12 tabs.

Goles, R.W.; Sevigny, G.J.; Andersen, C.M.

1990-06-01T23:59:59.000Z

138

Proceedings of the 17th Biennial Waste Processing Conference WASTE SEPARATION-  

E-Print Network (OSTI)

Proceedings of the 17th Biennial Waste Processing Conference ASME 1996 WASTE SEPARATION- DOES IT INFLUENCE MUNICIPAL WASTE COMBUSTOR EMISSIONS? A. John Chandler A.J. Chandler & Associates Ltd. Willowdale that MSW incinerator emissions show significant variations because of the heterogeneous nature of the waste

Columbia University

139

Process Waste Assessment for the Diana Laser Laboratory  

SciTech Connect

This Process Waste Assessment was conducted to evaluate the Diana Laser Laboratory, located in the Combustion Research Facility. It documents the hazardous chemical waste streams generated by the laser process and establishes a baseline for future waste minimization efforts. This Process Waste Assessment will be reevaluated in approximately 18 to 24 months, after enough time has passed to implement recommendations and to compare results with the baseline established in this assessment.

Phillips, N.M.

1993-12-01T23:59:59.000Z

140

Performance Evaluation of Advanced LLW Liquid Processing Technology: Boiling Water Reactor Liquid Processing  

Science Conference Proceedings (OSTI)

This report provides condensed information on boiling water reactor (BWR) membrane based liquid radwaste processing systems. The report presents specific details of the technology, including design, configuration, and performance. This information provides nuclear plant personnel with data useful in evaluating the merits of applying advanced processes at their plant.

2001-11-26T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Rotor for processing liquids using movable capillary tubes  

DOE Patents (OSTI)

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.

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

1989-05-30T23:59:59.000Z

142

Rotor for processing liquids using movable capillary tubes  

DOE Patents (OSTI)

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.

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

1987-07-17T23:59:59.000Z

143

Rotor for processing liquids using movable capillary tubes  

DOE Patents (OSTI)

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.

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

1989-01-01T23:59:59.000Z

144

Savannah River Site Marks Waste Processing Milestone with Melter's  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Marks Waste Processing Milestone with Marks Waste Processing Milestone with Melter's 2,000th Waste Canister Savannah River Site Marks Waste Processing Milestone with Melter's 2,000th Waste Canister February 1, 2012 - 12:00pm Addthis A Savannah River Remediation employee uses a manipulator located inside a shielded enclosure at the Defense Waste Processing Facility, where a melter pours molten glass into a canister. A Savannah River Remediation employee uses a manipulator located inside a shielded enclosure at the Defense Waste Processing Facility, where a melter pours molten glass into a canister. AIKEN, S.C. - The second melter to operate in the 16-year history of the nation's largest radioactive waste glassification plant shows no signs of slowing after recently pouring its 2,000 canister of glass-formed hazardous

145

Savannah River Site Marks Waste Processing Milestone with Melter's  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Savannah River Site Marks Waste Processing Milestone with Savannah River Site Marks Waste Processing Milestone with Melter's 2,000th Waste Canister Savannah River Site Marks Waste Processing Milestone with Melter's 2,000th Waste Canister February 1, 2012 - 12:00pm Addthis A Savannah River Remediation employee uses a manipulator located inside a shielded enclosure at the Defense Waste Processing Facility, where a melter pours molten glass into a canister. A Savannah River Remediation employee uses a manipulator located inside a shielded enclosure at the Defense Waste Processing Facility, where a melter pours molten glass into a canister. AIKEN, S.C. - The second melter to operate in the 16-year history of the nation's largest radioactive waste glassification plant shows no signs of slowing after recently pouring its 2,000 canister of glass-formed hazardous

146

Idaho Chemical Processing Plant low-activity waste grout stabilization development program FY-97 status report  

SciTech Connect

The general purpose of the Grout Development Program is to solidify and stabilize the liquid low-activity wastes (LAW) generated at the Idaho Chemical Processing Plant (ICPP). It is anticipated that LAW will be produced from the following: (1) chemical separation of the tank farm high-activity sodium-bearing waste, (2) retrieval, dissolution, and chemical separation of the aluminum, zirconium, and sodium calcines, (3) facility decontamination processes, and (4) process equipment waste. Grout formulation studies for sodium-bearing LAW, including decontamination and process equipment waste, continued this fiscal year. A second task was to develop a grout formulation to solidify potential process residual heels in the tank farm vessels when the vessels are closed.

Herbst, A.K.; Marshall, D.W.; McCray, J.A.

1998-02-01T23:59:59.000Z

147

Materials and Processes to Immobilize Nuclear Waste  

Science Conference Proceedings (OSTI)

Oct 8, 2012 ... While borosilicate glass is widely regarded as baseline technology for nuclear waste immobilisation, there are a wide range of such wastes that ...

148

Design of waste tyre pyrolysis process.  

E-Print Network (OSTI)

??xviii, 164 p. : ill. (some col.) ; 30 cm HKUST Call Number: Thesis CBME 2009 LeeK Waste tyre, one kind of non-biodegradable solid wastes,… (more)

Lee, King Lung

2009-01-01T23:59:59.000Z

149

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

SciTech Connect

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.

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

1993-03-01T23:59:59.000Z

150

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

Science Conference Proceedings (OSTI)

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

Not Available

1980-08-01T23:59:59.000Z

151

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

Energy.gov (U.S. Department of Energy (DOE))

This is a presentation outlining the Salt Waste Processing Facility process, major risks, approach for conducting reviews, discussion of the findings, and conclusions.

152

Analysis of Advanced Liquid Waste Minimization Techniques at a PWR: Advanced Media, Pleated Filters, and Ecomomic Evaluation Tools  

Science Conference Proceedings (OSTI)

Utilities may employ a number of options for processing radioactive liquids or improving processing system O&M. This report summarizes low level waste minimization studies for the Diablo Canyon Power Plant. These studies involved the performance of selective ion media, optimization of the chemical volume control system (CVCS) demineralizers, performance assessment of the application of advanced minimum precoat elements for processing condensate demineralizer system rinse water, and evaluation of the econ...

1998-06-30T23:59:59.000Z

153

Double Shell Tank (DST) Process Waste Sampling Subsystem Specification  

SciTech Connect

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.

RASMUSSEN, J.H.

2000-05-03T23:59:59.000Z

154

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

E-Print Network (OSTI)

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

Zhou, Yaoqi

155

GEOTECHNICAL/GEOCHEMICAL CHARACTERIZATION OF ADVANCED COAL PROCESS WASTE STREAMS  

Science Conference Proceedings (OSTI)

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.

Edwin S. Olson; Charles J. Moretti

1999-11-01T23:59:59.000Z

156

BLENDING ANALYSIS FOR RADIOACTIVE SALT WASTE PROCESSING FACILITY  

SciTech Connect

Savannah River National Laboratory (SRNL) evaluated methods to mix and blend the contents of the blend tanks to ensure the contents are properly blended before they are transferred from the blend tank such as Tank 21 and Tank 24 to the Salt Waste Processing Facility (SWPF) feed tank. The tank contents consist of three forms: dissolved salt solution, other waste salt solutions, and sludge containing settled solids. This paper focuses on developing the computational model and estimating the operation time of submersible slurry pump when the tank contents are adequately blended prior to their transfer to the SWPF facility. A three-dimensional computational fluid dynamics approach was taken by using the full scale configuration of SRS Type-IV tank, Tank 21H. Major solid obstructions such as the tank wall boundary, the transfer pump column, and three slurry pump housings including one active and two inactive pumps were included in the mixing performance model. Basic flow pattern results predicted by the computational model were benchmarked against the SRNL test results and literature data. Tank 21 is a waste tank that is used to prepare batches of salt feed for SWPF. The salt feed must be a homogeneous solution satisfying the acceptance criterion of the solids entrainment during transfer operation. The work scope described here consists of two modeling areas. They are the steady state flow pattern calculations before the addition of acid solution for tank blending operation and the transient mixing analysis during miscible liquid blending operation. The transient blending calculations were performed by using the 95% homogeneity criterion for the entire liquid domain of the tank. The initial conditions for the entire modeling domain were based on the steady-state flow pattern results with zero second phase concentration. The performance model was also benchmarked against the SRNL test results and literature data.

Lee, S.

2012-05-10T23:59:59.000Z

157

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

Science Conference Proceedings (OSTI)

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.

Smith, M.; Iverson, D.

2010-12-08T23:59:59.000Z

158

CRYSTALLINE CERAMIC WASTE FORMS: COMPARISON OF REFERENCE PROCESS FOR CERAMIC WASTE FORM FABRICATION  

SciTech Connect

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.

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

159

Construction Begins on New Waste Processing Facility | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Construction Begins on New Waste Processing Facility Construction Begins on New Waste Processing Facility Construction Begins on New Waste Processing Facility February 9, 2012 - 12:00pm Addthis Workers construct a new facility that will help Los Alamos National Laboratory accelerate the shipment of transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad for permanent disposal. Workers construct a new facility that will help Los Alamos National Laboratory accelerate the shipment of transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad for permanent disposal. Construction has begun on a new facility that will help Los Alamos National Laboratory accelerate the shipment of transuranic (TRU) waste stored in large boxes at Technical Area 54, Area G. Construction has begun on a new facility that will help Los Alamos National

160

Ultrafiltration treatment for liquid laundry wastes from nuclear power stations  

SciTech Connect

The authors conduct a comprehensive analysis of the waste constituents--radioactive and organic--of the laundry water resulting from the on-site laundering and decontamination of clothing worn in nuclear power plants. The primary isotope contaminants consist of niobium and zirconium 95, manganese 54, cobalt 60, iron 59, and cesium 134 and 137. A variety of filter and adsorbent materials used in an ultrafiltration process are comparatively tested for their effectiveness in removing not only these isotopes but also the organic contaminants in the process of recycling the water. Those materials consist of copper hexacyanoferrate, polyacrylophosphonic acid, and several metal-polymer complexes.

Kichik, V.A.; Maslova, M.N.; Svittsov, A.A.; Kuleshov, N.F.

1988-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

The Use of Transportable Processing Systems for the Treatment of Radioactive Nuclear Wastes  

Science Conference Proceedings (OSTI)

EnergySolutions has developed two major types of radioactive processing plants based on its experience in the USA and UK, and its exclusive North American access to the intellectual property and know-how developed over 50 years at the Sellafield nuclear site in the UK. Passive Secure Cells are a type of hot cell used in place of the Canyons typically used in US-designed radioactive facilities. They are used in permanent, large scale plants suitable for long term processing of large amounts of radioactive material. The more recently developed Transportable Processing Systems, which are the subject of this paper, are used for nuclear waste processing and clean-up when processing is expected to be complete within shorter timescales and when it is advantageous to be able to move the processing equipment amongst a series of geographically spread-out waste treatment sites. Such transportable systems avoid the construction of a monolithic waste processing plant which itself would require extensive decommissioning and clean-up when its mission is complete. This paper describes a range of transportable radioactive waste processing equipment that EnergySolutions and its partners have developed including: the portable MOSS drum-based waste grouting system, the skid mounted MILWPP large container waste grouting system, the IPAN skid-mounted waste fissile content non-destructive assay system, the Wiped Film Evaporator low liquid hold-up transportable evaporator system, the CCPU transportable solvent extraction cesium separation system, and the SEP mobile shielded cells for emptying radioactive debris from water-filled silos. Maximum use is made of proven, robust, and compact processing equipment such as centrifugal contactors, remote sampling systems, and cement grout feed and metering devices. Flexible, elastomer-based Hose-in-Hose assemblies and container-based transportable pump booster stations are used in conjunction with these transportable waste processing units for transferring radioactive waste from its source to the processing equipment. (authors)

Phillips, Ch.; Houghton, D.; Crawford, G. [EnergySolutions LLC., 2345 Stevens Drive, Richland, WA (United States)

2008-07-01T23:59:59.000Z

162

Process for preparing lubricating oil from used waste lubricating oil  

DOE Patents (OSTI)

A re-refining process is described by which high-quality finished lubricating oils are prepared from used waste lubricating and crankcase oils. The used oils are stripped of water and low-boiling contaminants by vacuum distillation and then dissolved in a solvent of 1-butanol, 2-propanol and methylethyl ketone, which precipitates a sludge containing most of the solid and liquid contaminants, unspent additives, and oxidation products present in the used oil. After separating the purified oil-solvent mixture from the sludge and recovering the solvent for recycling, the purified oil is preferably fractional vacuum-distilled, forming lubricating oil distillate fractions which are then decolorized and deodorized to prepare blending stocks. The blending stocks are blended to obtain a lubricating oil base of appropriate viscosity before being mixed with an appropriate additive package to form the finished lubricating oil product.

Whisman, Marvin L. (Bartlesville, OK); Reynolds, James W. (Bartlesville, OK); Goetzinger, John W. (Bartlesville, OK); Cotton, Faye O. (Bartlesville, OK)

1978-01-01T23:59:59.000Z

163

Independent Oversight Assessment, Salt Waste Processing Facility Project -  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Salt Waste Processing Facility Salt Waste Processing Facility Project - January 2013 Independent Oversight Assessment, 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 of Energy (DOE) Office of Enforcement and Oversight (Independent Oversight), within the Office of Health, Safety and Security (HSS), conducted an independent assessment of nuclear safety culture at the Salt Waste Processing Facility (SWPF) Project. The primary objective of the evaluation was to provide information regarding the status of the safety culture at the SWPF Project. The data collection phase of the assessment occurred during August - September 2012. Independent Oversight Assessment, Salt Waste Processing Facility Project -

164

Hydrothermal processing of Hanford tank wastes: Process modeling and control  

Science Conference Proceedings (OSTI)

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.

Currier, R.P. [comp.

1994-10-01T23:59:59.000Z

165

Status of Waste Processing Technology Development  

Radiation stability testing on sRF Evaluated and selected potential ... Technical reports Tests with real waste Program performance reviews

166

Process for treating fission waste. [Patent application  

DOE Patents (OSTI)

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.

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

1981-11-17T23:59:59.000Z

167

Melt-processed Multiphasic Ceramic Waste Forms  

Science Conference Proceedings (OSTI)

Symposium, Materials Issues in Nuclear Waste Management in the 21st Century ... Scanning electron microscopy (SEM) and energy dispersive spectrometry ...

168

Detection of free liquid in drums of radioactive waste. [Patent application  

DOE Patents (OSTI)

A nondestructive thermal imaging method for detecting the presence of a liquid such as water within a sealed container is described. The process includes application of a low amplitude heat pulse to an exterior surface area of the container, terminating the heat input and quickly mapping the resulting surface temperatures. The various mapped temperature values can be compared with those known to be normal for the container material and substances in contact. The mapped temperature values show up in different shades of light or darkness that denote different physical substances. The different substances can be determined by direct observation or by comparison with known standards. The method is particularly applicable to the detection of liquids above solidified radioactive wastes stored in sealed containers.

Not Available

1979-10-16T23:59:59.000Z

169

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

NLE Websites -- All DOE Office Websites (Extended Search)

River Site Salt Waste Processing Facility Safety Basis and Design Development May 2011 August 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement...

170

Waste Heat Recovery from Industrial Process Heating Equipment...  

NLE Websites -- All DOE Office Websites (Extended Search)

Waste Heat Recovery from Industrial Process Heating Equipment - Cross-cutting Research and Development Priorities Speaker(s): Sachin Nimbalkar Date: January 17, 2013 - 11:00am...

171

Process Chemistry and Operations Planning for Hanford Waste ...  

Process Chemistry and Operations Planning for Hanford Waste Alternatives L. T. Smith,* R. K. Toghiani, and J. S. Lindner Institute for Clean Energy Technology (ICET ...

172

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Radioactive Liquid Waste Radioactive Liquid Waste Treatment Facility Replacement Project at Los Alamos National Laboratory OAS-L-13-15 September 2013 Department of Energy Washington, DC 20585 September 26, 2013 MEMORANDUM FOR THE ASSOCIATE ADMINISTRATOR FOR ACQUISITION AND PROJECT MANAGEMENT MANAGER LOS ALAMOS FIELD OFFICE FROM: David Sedillo Western Audits Division Office of Inspector General SUBJECT: INFORMATION: Audit Report on "The Radioactive Liquid Waste Treatment Facility Replacement Project at Los Alamos National Laboratory" BACKGROUND The Department of Energy's Los Alamos National Laboratory (Los Alamos) is a Government- owned, contractor operated Laboratory that is part of the National Nuclear Security Administration's (NNSA) nuclear weapons complex. Los Alamos' primary responsibility is to

173

Thermocatalytic conversion of food processing wastes: Topical report, FY 1988  

DOE Green Energy (OSTI)

The efficient utilization of waste produced during food processing operations is a topic of growing importance to the industry. While incineration is an attractive option for wastes with relatively low ash and moisture contents (i.e., under about 50 wt % moisture), it is not suitable for wastes with high moisture contents. Cheese whey, brewer's spent grain, and fruit pomace are examples of food processing wastes that are generally too wet to burn efficiently and cleanly. Pacific Northwest Laboratory (PNL) is developing a thermocatalytic conversion process that can convert high-moisture wastes (up to 98 wt % moisture) to a medium-Btu fuel gas consisting primarily of methane and carbon dioxide. At the same time, the COD of these waste streams is reduced by 90% to 99%, Organic wastes are converted by thermocatalytic treatment at 350/degree/C to 400/degree/C and 3000 to 4000 psig. The process offers a relatively simple solution to waste treatment while providing net energy production from wastes containing as little as 2 wt % organic solids (this is equivalent to a COD of approximately 25,000 mg/L). This report describes continuous reactor system (CRS) experiments that have been conducted with food processing wastes. The purpose of the CRS experiments was to provide kinetic and catalyst lifetime data, which could not be obtained with the batch reactor tests. These data are needed for commercial scaleup of the process.

Baker, E.G.; Butner, R.S.; Sealock, L.J. Jr.; Elliott, D.C.; Neuenschwander, G.G.

1989-01-01T23:59:59.000Z

174

Mixed Waste Advanced Treatment Technology: Waste Processing Products and Their Recycling Applications  

Science Conference Proceedings (OSTI)

During their operations, nuclear power plants generate mixed waste containing both hazardous and radioactive constituents. Disposal options for such mixed waste are limited and expensive. EPRI research has demonstrated that an innovative molten metal process for destroying hazardous wastes can be used effectively on nuclear power plant wastes containing both hazardous and radioactive constituents. Preliminary results of this research indicate that the destruction of the hazardous constituents is complete...

1997-12-31T23:59:59.000Z

175

Thermal processing system concepts and considerations for RWMC buried waste  

SciTech Connect

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.

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

1992-02-01T23:59:59.000Z

176

Chapter 38 Hazardous Waste Permitting Process (Kentucky) | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

8 Hazardous Waste Permitting Process (Kentucky) 8 Hazardous Waste Permitting Process (Kentucky) Chapter 38 Hazardous Waste Permitting Process (Kentucky) < Back Eligibility Agricultural Commercial Construction Developer Fed. Government Industrial Institutional Investor-Owned Utility Local Government Municipal/Public Utility Rural Electric Cooperative Schools State/Provincial Govt Transportation Tribal Government Utility Savings Category Alternative Fuel Vehicles Hydrogen & Fuel Cells Program Info State Kentucky Program Type Environmental Regulations Provider Department for Environmental Protection This administrative regulation establishes the general provisions for storage, treatment, recycling, or disposal of hazardous waste. It provides information about permits and specific requirements for containers, tanks,

177

EVALUATION OF ULTIMATE DISPOSAL METHODS FOR LIQUID AND SOLID RADIOACTIVE WASTES. PART II. CONVERSION TO SOLID BY POT CALCINATION  

SciTech Connect

The costs of pot calcination of Purex and Thorex wastes were calculated. The wastes were assumed produced by a plant processing 1500 ton/year of U converter fuel at a burnup of 10,000 Mwd/ton and 270 ton/year of Th converter fuel at 20,000 Mwd/ton. Costs were calculated for processing Purex waste in acidic and reacidified forms and for processing Thorex wastes in acidic and reacidified forms and with constituents added for producing an acidic Thorex glass. Calcination vessel designs were right circular cylinders similar to those used in engineering development studies. Costs were calculated for processing in 6-, 12-, and 24-in.-dia vessels with a fixed length of 10 ft. Vessel costs used, based on estimates from private industry, were calculated for wastes decayed 120 days and 1, 3, 10, and 30 years after reactor discharge prior to calcination. Aging had negligible effect on costs, except as it permitted larger diameter vessels to be used, because vessel and operating costs were much larger than capital costs in all cases. The lowest cost was 0.87 x 10/sup -2/ mill/kwh/sub e/ for processing acidic Purex and Thorex wastes in 24-in.-dia vessels, and the highest was 5.0 x 10/sup -2/ mill/kwh/sub e/ for processing reacidified Purex and Thorex wastes in 6-in.-dia vessels. About 7 years of interim liquid storage would be required before acidic Purex wastes could be processed in 24-in.-dia vessels. (auth)

Perona, J.J.; Bradshaw, R.L.; Roberts, J.T.; Blomeke, J.O.

1961-10-16T23:59:59.000Z

178

Sodium Bearing Waste Processing Alternatives Analysis  

SciTech Connect

A multidisciplinary team gathered to develop a BBWI recommendation to DOE-ID on the processing alternatives for the sodium bearing waste in the INTEC Tank Farm. Numerous alternatives were analyzed using a rigorous, systematic approach. The data gathered were evaluated through internal and external peer reviews for consistency and validity. Three alternatives were identified to be top performers: Risk-based Calcination, MACT to WIPP Calcination and Cesium Ion Exchange. A dual-path through early Conceptual design is recommended for MACT to WIPP Calcination and Cesium Ion Exchange since Risk-based Calcination does not require design. If calcination alternatives are not considered based on giving Type of Processing criteria significantly greater weight, the CsIX/TRUEX alternative follows CsIX in ranking. However, since CsIX/TRUEX shares common uncertainties with CsIX, reasonable backups, which follow in ranking, are the TRUEX and UNEX alternatives. Key uncertainties must be evaluated by the decision-makers to choose one final alternative. Those key uncertainties and a path forward for the technology roadmapping of these alternatives is provided.

Murphy, James Anthony; Palmer, Brent J; Perry, Keith Joseph

2003-12-01T23:59:59.000Z

179

Radioactive Liquid Waste Treatment Facility Discharges in 2011  

Science Conference Proceedings (OSTI)

This report documents radioactive discharges from the TA50 Radioactive Liquid Waste Treatment Facilities (RLWTF) during calendar 2011. During 2011, three pathways were available for the discharge of treated water to the environment: discharge as water through NPDES Outfall 051 into Mortandad Canyon, evaporation via the TA50 cooling towers, and evaporation using the newly-installed natural-gas effluent evaporator at TA50. Only one of these pathways was used; all treated water (3,352,890 liters) was fed to the effluent evaporator. The quality of treated water was established by collecting a weekly grab sample of water being fed to the effluent evaporator. Forty weekly samples were collected; each was analyzed for gross alpha, gross beta, and tritium. Weekly samples were also composited at the end of each month. These flow-weighted composite samples were then analyzed for 37 radioisotopes: nine alpha-emitting isotopes, 27 beta emitters, and tritium. These monthly analyses were used to estimate the radioactive content of treated water fed to the effluent evaporator. Table 1 summarizes this information. The concentrations and quantities of radioactivity in Table 1 are for treated water fed to the evaporator. Amounts of radioactivity discharged to the environment through the evaporator stack were likely smaller since only entrained materials would exit via the evaporator stack.

Del Signore, John C. [Los Alamos National Laboratory

2012-05-16T23:59:59.000Z

180

Waste Processing and Recycling: Some Case Studies  

Science Conference Proceedings (OSTI)

Symposium, WASTE RECYCLING IN MINERAL AND METALLURGICAL ... Effect of Electricity Mix and Ore Grade on the Carbon Footprint of Chilean Cathodic ...

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181

DOE Selects Savannah River Remediation, LLC for Liquid Waste Contract at  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOE Selects Savannah River Remediation, LLC for Liquid Waste DOE Selects Savannah River Remediation, LLC for Liquid Waste Contract at Savannah River Site DOE Selects Savannah River Remediation, LLC for Liquid Waste Contract at Savannah River Site December 8, 2008 - 4:58pm Addthis Washington, D.C. -The U.S. Department of Energy (DOE) today announced the award to Savannah River Remediation, LLC as the liquid waste contractor for DOE's Savannah River Site (SRS) in Aiken, South Carolina. The contract is a cost-plus award-fee contract valued at approximately $3.3 billion over the entire contract, consisting of a base period of six years, plus an option to extend for up to two additional years. The base performance period of the contract will be from April 1, 2009 through March 31, 2015. A 90-day transition period will begin January 2, 2009.

182

DOE Selects Savannah River Remediation, LLC for Liquid Waste Contract at  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOE Selects Savannah River Remediation, LLC for Liquid Waste DOE Selects Savannah River Remediation, LLC for Liquid Waste Contract at Savannah River Site DOE Selects Savannah River Remediation, LLC for Liquid Waste Contract at Savannah River Site December 8, 2008 - 4:58pm Addthis Washington, D.C. -The U.S. Department of Energy (DOE) today announced the award to Savannah River Remediation, LLC as the liquid waste contractor for DOE's Savannah River Site (SRS) in Aiken, South Carolina. The contract is a cost-plus award-fee contract valued at approximately $3.3 billion over the entire contract, consisting of a base period of six years, plus an option to extend for up to two additional years. The base performance period of the contract will be from April 1, 2009 through March 31, 2015. A 90-day transition period will begin January 2, 2009.

183

Use of the Environmental Simulation Program (ESP) to Simulate Complex Waste Treatment Processes  

Science Conference Proceedings (OSTI)

The Environmental Simulation Program is a process simulator designed for aqueous based chemical processes. ESP, which is produced by OLI Systems, Inc., utilizes sophisticated activity coefficient models and predictive equations that result in the ability to simulate very complex electrolyte systems (OLI, 2002). The software comes with databanks of regressed parameters for a large number of aqueous, vapor, and solid species covering most of the elements. ESP has been used extensively at the U. S. Department of Energy Hanford Site to predict nuclear waste slurry vapor-liquid-solid equilibrium. It has and is being used to model leaching and washing of nuclear waste sludges, evaporation of nuclear waste solutions, crystallization of salts, precipitation of plutonium and other metals from waste solutions, and other processing of dilute and concentrated aqueous solutions, sludges, and slurries. The software is also used extensively to rationalize the characterization of nuclear wastes using limited data from analyses of waste samples. The OLI provided databanks suffer from a legacy interaction model that limits the accuracy when neutral solutes are important. Also, the nitrate-nitrite systems typically found in nuclear wastes are not properly parameterized in ESP databases because of the existence of sodium nitrate and nitrite ion pairs. Properties databanks for ESP have been developed at Flour Federal Services that eliminate the legacy model and provide more accurate simulation results than the OLI supplied databases for such concentrated solutions and slurries.

MacLean, G. T.; Ho, Q. T.; Berger, S. R. K.

2003-02-26T23:59:59.000Z

184

Independent Oversight Review, Oak Ridge Transuranic Waste Processing  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Oak Ridge Transuranic Waste Oak Ridge Transuranic Waste Processing Center, September 2013 Independent Oversight Review, Oak Ridge Transuranic Waste Processing Center, September 2013 September 2013 Review of Management 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 oversight review of the management of safety significant structures, systems, and components at the Oak Ridge Transuranic Waste Processing Center (TWPC). The review was performed April 2-5, April 15-19, and May 19-23, 2013, by the Department of Energy's (DOE) Office of Safety and Emergency Management Evaluations, which is within the DOE Office of Health, Safety and Security. The review was carried out within the broader context of an ongoing program of

185

Liquid low-level waste generation projections for ORNL in 1993  

SciTech Connect

Liquid low-level waste (LLLW) is generated by various programs and projects throughout Oak Ridge National Laboratory (ORNL). These wastes are collected in underground collection tanks, bottles, and trucks; they are then neutralized with sodium hydroxide and treated for volume reduction at the ORNL evaporator facility. This report presents historical and projected data concerning the volume and characterization of LLLW, prior to and after evaporation. Storage space for projected waste generation is also discussed.

DePaoli, S.M.

1994-04-01T23:59:59.000Z

186

Double liquid membrane system for the removal of actinides and lanthanides from acidic nuclear wastes  

SciTech Connect

Supported liquid membranes (SLM), consisting of an organic solution of n-octyl-(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) and tributyl-phosphate (TBP) in decalin are able to perform selective separation and concentration of actinide and lanthanide ions from aqueous nitrate feed solutions and synthetic nuclear wastes. In the membrane process a possible strip solution is a mixture of formic acid and hydroxylammonium formate (HAF). The effectiveness of this strip solution is reduced and eventually nullified by the simultaneous transfer through the SLM of nitric acid which accumulates in the strip solution. A possible way to overcome this drawback is to make use of a second SLM consisting of a primary amine which is able to extract only HNO/sub 3/ from the strip solution. In this work the results obtained by experimentally studying the membrane system: synthetic nuclear waste/CMPO-TBP membrane/HCOOH-HAF strip solution/primary amine membrane/NaOH solution, are reported. They show that the use of a second liquid membrane is effective in controlling the HNO/sub 3/ concentration in the strip solution, thus allowing the actinide and lanthanide ions removal from the feed solution to proceed to completion. 15 refs., 10 figs., 1 tab.

Chiarizia, R.; Danesi, P.R.

1985-01-01T23:59:59.000Z

187

A comparision of TRUEX and CMP solvent extraction processes for actinide removal from ICPP wastes  

SciTech Connect

The Idaho Chemical Processing Plant (ICPP) is currently engaged in development efforts for the decontamination of high-level radioactive wastes generated from decades of nuclear fuel reprocessing. These wastes include several types of calcine, generated by high temperature solidification of reprocessing raffinates. In addition to calcine, there are smaller quantities of secondary wastes from decontamination and solvent wash activities which are typically referred to as sodium-bearing waste (SBW). Solvent extraction technologies based on octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide (CMPO, the active extractant in the TRUEX process) and dihexyl-N,N-diethylcarbamoylmethylphosphonate (DHDECMP, the active extractant in the CMP process) are being evaluated for actinide partitioning from these waste streams. Calcines must first be dissolved in an appropriate acidic solution prior to treatment in solvent extraction based processes. The SBW is currently stored as an acidic solution and readily amenable to liquid extraction techniques. Development efforts to date have revolved around defining and refining baseline flowsheets with the TRUEX and CMP processes for each waste stream. Another objective of this work was to determine which of these technologies are best suited for the treatment of ICPP wastes. Laboratory batch contacts were performed to identify relevant chemistry and distribution coefficients. This information was then used to establish baseline flowsheet configuration with regard to chemistry. The laboratory data were used to model the behavior of the actinides and other constituents in the wastes in countercurrent, continuous processes based on centrifugal contactor technology. The laboratory data and modelling results form the basis for comparison of the two processes.

Herbst, R.S.; Brewer, K.N.; Garn, T.G.; Law, J.D. [and others

1996-04-01T23:59:59.000Z

188

Independent Oversight Review, Savannah River Site Salt Waste Processing  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Savannah River Site Salt Waste Savannah River Site Salt Waste Processing Facility - August 2013 Independent Oversight Review, Savannah River Site Salt Waste Processing Facility - August 2013 August 2013 Review of the Savannah River Site Salt Waste Processing Facility Safety Basis and Design Development. This report documents the results of an independent oversight review of the safety basis and design development for the Salt Waste Processing Facility (SWPF) at the U.S. Department of Energy (DOE) Savannah River Site. The review was performed February 12-14, 2013 by DOE's Office of Safety and Emergency Management Evaluations, which is within the DOE Office of Health, Safety and Security. The purpose of the review was to evaluate the safety basis, design, and the associated technical documents developed for

189

Independent Oversight Review, Savannah River Site Salt Waste Processing  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Savannah River Site Salt Waste Savannah River Site Salt Waste Processing Facility - August 2013 Independent Oversight Review, Savannah River Site Salt Waste Processing Facility - August 2013 August 2013 Review of the Savannah River Site Salt Waste Processing Facility Safety Basis and Design Development. This report documents the results of an independent oversight review of the safety basis and design development for the Salt Waste Processing Facility (SWPF) at the U.S. Department of Energy (DOE) Savannah River Site. The review was performed February 12-14, 2013 by DOE's Office of Safety and Emergency Management Evaluations, which is within the DOE Office of Health, Safety and Security. The purpose of the review was to evaluate the safety basis, design, and the associated technical documents developed for

190

Independent Oversight Review, Oak Ridge Transuranic Waste Processing  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Oak Ridge Transuranic Waste Oak Ridge Transuranic Waste Processing Facility - December 2013 Independent Oversight Review, Oak Ridge Transuranic Waste Processing Facility - December 2013 December 2013 Review of the Fire Protection Program and Fire Protection Systems at the Transuranic Waste Processing Center This report documents the results of an independent oversight review of the fire protection programs and systems at the Oak Ridge Transuranic Waste Processing Center. The review was performed during May 20-23, 2013, and July 15-19, 2013, by the U.S. Department of Energy's (DOE) Office of Safety and Emergency Management Evaluations, which is within the DOE Office of Health, Safety and Security. The review was one part of a targeted assessment of fire protection at nuclear facilities across the DOE complex.

191

Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-98 Status Report  

SciTech Connect

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 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

Herbst, Alan Keith; Mc Cray, John Alan; Rogers, Adam Zachary; Simmons, R. F.; Palethorpe, S. J.

1999-03-01T23:59:59.000Z

192

Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program, FY-98 Status Report  

SciTech Connect

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 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

Herbst, A.K.; Rogers, A.Z.; McCray, J.A.; Simmons, R.F.; Palethorpe, S.J.

1999-03-01T23:59:59.000Z

193

High level radioactive waste vitrification process equipment component testing  

Science Conference Proceedings (OSTI)

Remote operability and maintainability of vitrification equipment were assessed under shielded-cell conditions. The equipment tested will be applied to immobilize high-level and transuranic liquid waste slurries that resulted from plutonium production for defense weapons. Equipment tested included: a turntable for handling waste canisters under the melter; a removable discharge cone in the melter overflow section; a thermocouple jumper that extends into a shielded cell; remote instrument and electrical connectors; remote, mechanical, and heat transfer aspects of the melter glass overflow section; a reamer to clean out plugged nozzles in the melter top; a closed circuit camera to view the melter interior; and a device to retrieve samples of the glass product. A test was also conducted to evaluate liquid metals for use in a liquid metal sealing system.

Siemens, D.H.; Heath, W.O.; Larson, D.E.; Craig, S.N.; Berger, D.N.; Goles, R.W.

1985-04-01T23:59:59.000Z

194

Process for stabilization of coal liquid fractions  

SciTech Connect

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.

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

1987-01-01T23:59:59.000Z

195

Application of thermogravimetric analysis to study the thermal degradation of solid and liquid organic wastes  

Science Conference Proceedings (OSTI)

In this work, the thermolysis of composite binary mixtures of refinery or coal-processing waste with waste biomass and D-grade (long-flame) coal was analyzed in order to increase the efficiency of the cothermolysis of chemically different organic wastes mainly because of the synergism of the thermolysis of mixture components and, correspondingly, the selectivity of formation of high-quality by-products (solid, gaseous, or liquid). A new approach to the analysis of thermogravimetric data was proposed and developed as applied to complex binary mixtures of carbon-containing materials. This approach was based on (1) the preliminary separation of the thermal degradation of individual carbon-containing mixture components into individual structural constituents and (2) the monitoring of the conversion of each particular structure fragment as a constituent of the mixtures in the course of the cothermolysis of the mixtures of starting components. Based on the approach developed, data on the main synergism effects in the course of cothermolysis in the binary test systems were obtained: the temperature regions of the appearance of these effects were distinguished, the main conclusions were made with respect to particular structure fragments in complex organic wastes responsible for the interaction of components in composite systems, and the directions (positive or negative) of changes in the yields of solid by-products and the degrees of effects (difference between the yields of cothermolysis by-products in each particular region of the appearance of synergistic effects in the systems) were determined. Additionally, the influence of alkali metal carbonate additives on synergistic effects in the interaction between binary system components under the process conditions of cothermolysis was analyzed.

E.S. Lygina; A.F. Dmitruk; S.B. Lyubchik; V.F. Tret'yakov [Tugan-Baranovsky State University of Economy and Trade, Donetsk (Ukraine)

2009-07-01T23:59:59.000Z

196

Biological Information Document, Radioactive Liquid Waste Treatment Facility  

SciTech Connect

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.

Biggs, J.

1995-12-31T23:59:59.000Z

197

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

Science Conference Proceedings (OSTI)

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.

RASMUSSEN, J.H.

2000-04-25T23:59:59.000Z

198

Idaho Chemical Processing Plant low-level waste grout stabilization development program FY-96 status report  

Science Conference Proceedings (OSTI)

The general purpose of the Grout Stabilization Development Program is to solidify and stabilize the liquid low-level wastes (LLW) generated at the Idaho Chemical Processing Plant (ICPP). It is anticipated that LLW will be produced from the following: (1) chemical separation of the tank farm high-activity sodium-bearing waste; (2) retrieval, dissolution, and chemical separation of the aluminum, zirconium, and sodium calcines; (3) facility decontamination processes; and (4) process equipment waste. The main tasks completed this fiscal year as part of the program were chromium stabilization study for sodium-bearing waste and stabilization and solidification of LLW from aluminum and zirconium calcines. The projected LLW will be highly acidic and contain high amounts of nitrates. Both of these are detrimental to Portland cement chemistry; thus, methods to precondition the LLW and to cure the grout were explored. A thermal calcination process, called denitration, was developed to solidify the waste and destroy the nitrates. A three-way blend of Portland cement, blast furnace slag, and fly ash was successfully tested. Grout cubes were prepared at various waste loadings to maximize loading while meeting compressive strength and leach resistance requirements. For the sodium LLW, a 25% waste loading achieves a volume reduction of 3.5 and a compressive strength of 2,500 pounds per square inch while meeting leach, mix, and flow requirements. It was found that the sulfur in the slag reduces the chromium leach rate below regulatory limits. For the aluminum LLW, a 15% waste loading achieves a volume reduction of 8.5 and a compressive strength of 4,350 pounds per square inch while meeting leach requirements. Likewise for zirconium LLW, a 30% waste loading achieves a volume reduction of 8.3 and a compressive strength of 3,570 pounds per square inch.

Herbst, A.K.

1996-09-01T23:59:59.000Z

199

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

SciTech Connect

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.

Van Hoesen, S.D.

2003-09-09T23:59:59.000Z

200

System for removing liquid waste from a tank  

DOE Patents (OSTI)

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.

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

1994-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

System for removing liquid waste from a tank  

DOE Patents (OSTI)

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.

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

1994-04-26T23:59:59.000Z

202

HANFORD'S SIMULATED LOW ACTIVITY WASTE CAST STONE PROCESSING  

SciTech Connect

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.

Kim, Y.

2013-08-20T23:59:59.000Z

203

Recovery of Tritium from Pharmaceutical Mixed Waste Liquids  

Science Conference Proceedings (OSTI)

Decontamination and Waste / Proceedings of the Sixth International Conference on Tritium Science and Technology Tsukuba, Japan November 12-16, 2001

W. T. Shmayda; R. D. Gallagher

204

Microsoft Word - FINAL 7-12-10 Site Visit Report - LANL Radioactive Liquid Waste Facility FCA.docx  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Site Visit Report Facility Centered Assessment of the Los Alamos National Laboratory Radioactive Liquid Waste Treatment Facility - June 2010 This site visit report documents the results of the Office of Health, Safety and Security's (HSS) review of the Facility Centered Assessment (FCA) of the Los Alamos National Laboratory (LANL) Radioactive Liquid Waste Treatment Facility (RLW). This review, conducted June 9-25, 2010, was sponsored by the U.S. Department of Energy (DOE) Los Alamos Site Office (LASO) and LANL, and conducted jointly by HSS, LASO, and LANL staff. The Office of Environment, Safety and Health Evaluations was the overall lead organization for evaluation of the FCA process with the participation of the LASO Facility Representative assigned to RLW.

205

Technical resource document for assured thermal processing of wastes  

Science Conference Proceedings (OSTI)

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.

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

1994-06-01T23:59:59.000Z

206

Waste heat driven absorption refrigeration process and system  

DOE Patents (OSTI)

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.

Wilkinson, William H. (Columbus, OH)

1982-01-01T23:59:59.000Z

207

Technology development program for Idaho Chemical Processing Plant spent fuel and waste management  

SciTech Connect

Irradiated nuclear fuel has been reprocessed at the Idaho Chemical Processing Plant (ICPP) since 1953 to recover uranium-235 and krypton-85 for the US Department of Energy (DOE). The resulting acidic high-level liquid radioactive waste (HLLW) has been solidified to a high-level waste (HLW) calcine since 1963 and stored in stainless-steel bins enclosed in concrete vaults. Residual HLW and radioactive sodium-bearing waste are stored in stainless-steel underground tanks contained in concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also stored at INEL. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium. As a result of the decision to curtail reprocessing the ICPP Spent Fuel and Waste Management Technology Development plan has been implemented to identify acceptable options for disposing of the (1) sodium-bearing liquid radioactive waste, (2) radioactive calcine, and (3) irradiated spent fuel stored at the INEL. The plan was developed jointly by DOE and Westinghouse Idaho Nuclear Company, Inc., (WINCO) and with the concurrence of the State of Idaho.

Ermold, L.F.; Knecht, D.A.; Hogg, G.W.; Olson, A.L.

1993-06-01T23:59:59.000Z

208

Accelerator Production of Tritium project process waste assessment  

Science Conference Proceedings (OSTI)

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).

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

1995-09-01T23:59:59.000Z

209

Catalytic hydrogenation process and apparatus with improved vapor liquid separation  

DOE Patents (OSTI)

A continuous hydrogenation process and apparatus wherein liquids are contacted with hydrogen in an ebullated catalyst reaction zone with the liquids and gas flowing vertically upwardly through that zone into a second zone substantially free of catalyst particles and wherein the liquid and gases are directed against an upwardly inclining surface through which vertical conduits are placed having inlet ends at different levels in the liquid and having outlet ends at different levels above the inclined surface, such that vapor-rich liquid is collected and discharged through conduits terminating at a higher level above the inclined surface than the vapor-poor liquid which is collected and discharged at a level lower than the inclined surface.

Chervenak, Michael C. (Pennington, NJ); Comolli, Alfred G. (Trenton, NJ)

1980-01-01T23:59:59.000Z

210

Recycling policy making of organic waste using analytical network process  

Science Conference Proceedings (OSTI)

The Analytic Hierarchy Process (AHP) has been used widely in multicriteria selection problems. However, AHP can deal with only a simple hierarchy of elements. On the other hand, the Analytical Network Process (ANP) can deal with more complex structures ... Keywords: analytical network process (ANP), group discussion, multicriteria selection, organic waste recycling policy making

Kazuei Ishii; Toru Furuichi

2008-11-01T23:59:59.000Z

211

Reliability analysis of common hazardous waste treatment processes  

Science Conference Proceedings (OSTI)

Five hazardous waste treatment processes are analyzed probabilistically using Monte Carlo simulation to elucidate the relationships between process safety factors and reliability levels. The treatment processes evaluated are packed tower aeration, reverse osmosis, activated sludge, upflow anaerobic sludge blanket, and activated carbon adsorption.

Waters, R.D. [Vanderbilt Univ., Nashville, TN (United States)

1993-05-01T23:59:59.000Z

212

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

SciTech Connect

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

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

2008-01-15T23:59:59.000Z

213

Zone Freezing Study for Pyrochemical Process Waste Minimization  

Science Conference Proceedings (OSTI)

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.

Ammon Williams

2012-05-01T23:59:59.000Z

214

Process for recovery of liquid hydrocarbons  

SciTech Connect

Methane is recovered as a gas for discharge to a pipeline from a gas stream containing methane and heavier hydrocarbons, principally ethane and propane. Separation is accomplished by condensing the heavier hydrocarbons and distilling the methane therefrom. A liquid product (LPG) comprising the heavier hydrocarbons is subsequently recovered and transferred to storage. Prior to being discharged to a pipeline, the recovered methane gas is compressed and in undergoing compression the gas is heated. The heat content of the gas is employed to reboil the refrigerant in an absorption refrigeration unit. The refrigeration unit is used to cool the LPG prior to its storage.

Millar, J.F.; Cockshott, J.E.

1978-04-11T23:59:59.000Z

215

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

SciTech Connect

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.

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

2011-11-07T23:59:59.000Z

216

DOE Office of Waste Processing Technical Exchange - Agenda  

NLE Websites -- All DOE Office Websites (Extended Search)

June 1, 2009 June 1, 2009 Agenda Hotel Information Registration Presentation Guidelines Poster Guidelines Webcast Waiver Contacts Home Waste Processing Technical Exchange Agenda (Version 1.1) Pre-Registration: Monday, May 18, 5:00p - 7:00p Organizer/Session Chair: Blocker (early registration & speaker check-in) Day 1: Tuesday, May 19 Registration - 7:00a - 8:00a Session One - Opening Session Two - Waste Retrieval and Closure 1 Session Three - Waste Form Development Day 2: Wednesday, May 20 Session Four - Pretreatment 1 Session Five - Facility Readiness and Start-up Session Six - Pretreatment 2 Session Seven - Waste Retrieval and Closure 2 Session Eight - Poster Presentations Day 3: Thursday, May 21 Session Nine - Regulatory Activity and Performance Assessment Session Ten - Waste Storage and Tank Farm Operational Improvements

217

WASTE TREATMENT TECHNOLOGY PROCESS DEVELOPMENT PLAN FOR HANFORD WASTE TREATMENT PLANT LOW ACTIVITY WASTE RECYCLE  

SciTech Connect

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 evalua

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

2013-08-29T23:59:59.000Z

218

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

SciTech Connect

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.

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

219

Novel Solvent for the Simultaneous recovery of Radioactive Nuclides from Liquid Radioactive Wastes  

DOE Patents (OSTI)

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.

Romanovskiy, Valeriy Nicholiavich; Smirnov, Lgor V.; Babain, Vasiliy A.; Todd, Terry A.; Brewer, Ken N.

1999-10-07T23:59:59.000Z

220

Process chemistry for the pretreatment of Hanford tank wastes  

SciTech Connect

Current guidelines for disposing radioactive wastes stored in underground tanks at the US Department of Energy`s Hanford Site call for the vitrification of high-level waste in borosilicate glass and disposal of the glass canisters in a deep geologic repository. Low-level waste is to be cast in grout and disposed of on site in shallow burial vaults. Because of the high cost of vitrification and geologic disposal, methods are currently being developed to minimize the volume of high-level waste requiring disposal. Two approaches are being considered for pretreating radioactive tank sludges: (1) leaching of selected components from the sludge and (2) acid dissolution of the sludge followed by separation of key radionuclides. The leaching approach offers the advantage of simplicity, but the acid dissolution/radionuclide extraction approach has the potential to produce the least number of glass canisters. Four critical components (Cr, P, S, and Al) were leached from an actual Hanford tank waste-Plutonium Finishing Plant sludge. The Al, P, and S were removed from the sludge by digestion of the sludge with 0.1 M NaOH at 100{degrees}C. The Cr was leached by treating the sludge with alkaline KMnO{sub 4} at 100{degrees}C. Removing these four components from the sludge will dramatically lower the number of glass canisters required to dispose of this waste. The transuranic extraction (TRUEX) solvent extraction process has been demonstrated at a bench scale using an actual Hanford tank waste. The process, which involves extraction of the transuranic elements with octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO), separated 99.9% of the transuranic elements from the bulk components of the waste. Several problems associated with the TRUEX processing of this waste have been addressed and solved.

Lumetta, G.J.; Swanson, J.L. [Pacific Northwest Lab., Richland, WA (United States); Barker, S.A. [Westinghouse Hanford Co., Richland, WA (United States)

1992-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

The Defense Waste Processing Facility: Two Years of Radioactive Operation  

Science Conference Proceedings (OSTI)

The Defense Waste Processing Facility (DWPF) at the Savannah River Site in Aiken, SC is currently immobilizing high level radioactive sludge waste in borosilicate glass. The DWPF began vitrification of radioactive waste in May, 1996. Prior to that time, an extensive startup test program was completed with simulated waste. The DWPF is a first of its kind facility. The experience gained and data collected during the startup program and early years of operation can provide valuable information to other similar facilities. This experience involves many areas such as process enhancements, analytical improvements, glass pouring issues, and documentation/data collection and tracking. A summary of this experience and the results of the first two years of operation will be presented.

Marra, S.L. [Westinghouse Savannah River Company, AIKEN, SC (United States); Gee, J.T.; Sproull, J.F.

1998-05-01T23:59:59.000Z

222

Simulation of waste processing, transportation, and disposal operations  

E-Print Network (OSTI)

In response to the accelerated cleanup goals of the Department of Energy, Sandia National Laboratory (Sandia) has developed and utilized a number of simulation models to represent the processing, transportation, and disposal of radioactive waste. Sandia, in conjunction with Simulation Dynamics, has developed a Supply Chain model of the cradle to grave management of radioactive waste. Sandia has used this model to assist the Department of Energy in developing a cost effective, regulatory compliant and efficient approach to dispose of waste from 25 sites across the country over the next 35 years. 1

Janis Trone

2000-01-01T23:59:59.000Z

223

Simulation Of Waste Processing, Transportation, And Disposal Operations  

E-Print Network (OSTI)

In response to the accelerated cleanup goals of the Department of Energy, Sandia National Laboratory (Sandia) has developed and utilized a number of simulation models to represent the processing, transportation, and disposal of radioactive waste. Sandia, in conjunction with Simulation Dynamics, has developed a Supply Chain model of the cradle to grave management of radioactive waste. Sandia has used this model to assist the Department of Energy in developing a cost effective, regulatory compliant and efficient approach to dispose of waste from 25 sites across the country over the next 35 years.

J. A. Joines; R. R. Barton; K. Kang; P. A. Fishwick; Janis Trone; Angela Guerin

2000-01-01T23:59:59.000Z

224

Process to separate transuranic elements from nuclear waste  

DOE Patents (OSTI)

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.

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

1989-03-21T23:59:59.000Z

225

Process to separate transuranic elements from nuclear waste  

DOE Patents (OSTI)

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.

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

1988-07-12T23:59:59.000Z

226

Improved Process control of wood waste fired boilers  

DOE Green Energy (OSTI)

This project's principal aim was the conceptual and feasibility stage development of improved process control methods for wood-waste-fired water-tube boilers operating in industrial manufacturing applications (primarily pulp and paper). The specific objectives put forth in the original project proposal were as follows: (1) fully characterize the wood-waste boiler control inter-relationships and constraints through data collection and analysis; (2) design an improved control architecture; (3) develop and test an appropriate control and optimization algorithm; and (4) develop and test a procedure for reproducing the approach and deriving the benefits on similar pulp and paper wood-waste boilers. Detailed tasks were developed supporting these objectives.

Process Control Solutions, Inc.

2004-01-30T23:59:59.000Z

227

Idaho Chemical Processing Plant Spent Fuel and Waste Management Technology Development Program Plan  

SciTech Connect

The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage and reprocessing since 1953. Reprocessing of SNF has resulted in an existing inventory of 1.5 million gallons of radioactive sodium-bearing liquid waste and 3800 cubic meters (m{sup 3}) of calcine, in addition to the 768 metric tons (MT) of SNF and various other fuel materials in inventory. To date, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, recent changes in world events have diminished the demand to recover and recycle this material. As a result, DOE has discontinued reprocessing SNF for uranium recovery, making the need to properly manage and dispose of these and future materials a high priority. In accordance with the Nuclear Waste Policy Act (NWPA) of 1982, as amended, disposal of SNF and high-level waste (HLW) is planned for a geological repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP Spent Fuel and Waste Management Technology Development Program (SF&WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will properly stored and prepared for final disposal. Program elements in support of acceptable interim storage and waste minimization include: developing and implementing improved radioactive waste treatment technologies; identifying and implementing enhanced decontamination and decommissioning techniques; developing radioactive scrap metal (RSM) recycle capabilities; and developing and implementing improved technologies for the interim storage of SNF.

1993-09-01T23:59:59.000Z

228

Conversion of historic waste treatment process for production of an LDR and WIPP/WAC compliant TRU wasteform  

SciTech Connect

In support of the historic weapons production mission at the, Rocky Flats Environmental Technology Site (RFETS), several liquid waste treatment processes were designed, built and operated for treatment of plutonium-contaminated aqueous waste. Most of these @ processes ultimately resulted in the production of a cemented wasteform. One of these treatment processes was the Miscellaneous Aqueous Waste Handling and Solidification Process, commonly referred to as the Bottlebox process. Due to a lack of processing demand, Bottlebox operations were curtailed in late 1989. Starting in 1992, a treatment capability for stabilization of miscellaneous, Resource Conservation and Recovery Act (RCRA) hazardous, plutonium-nitrate solutions was identified. This treatment was required to address potentially unsafe storage conditions for these liquids. The treatment would produce a TRU wasteform. It thus became necessary to restart the Bottlebox process, but under vastly different conditions and constraints than existed prior to its curtailment. This paper provides a description of the historical Bottlebox process and process controls; and then describes, in detail, all of the process and process control changes that were implemented to convert the treatment system such that a Waste Isolation Pilot Plant (WIPP) and a Land Disposal Requirements (LDR) compliant wasteform would be produced. The rationale for imposition of LDRs on a TRU wasteform is discussed. In addition, this paper discusses the program changes implemented to meet modem criticality safety, Conduct of Operations, and Department of Energy Nuclear Facility restart requirements.

Dunn, R.P.; Wagner, R.A.

1997-03-01T23:59:59.000Z

229

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

SciTech Connect

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

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

2002-01-17T23:59:59.000Z

230

Update of the management strategy for Oak Ridge National Laboratory Liquid Low-Level Waste  

Science Conference Proceedings (OSTI)

The strategy for management of the Oak Ridge National Laboratory`s (ORNL) radioactively contaminated liquid waste was reviewed in 1991. The latest information available through the end of 1990 on waste characterization, regulations, US Department of Energy (DOE) budget guidance, and research and development programs was evaluated to determine how the strategy should be revised. Few changes are needed to update the strategy to reflect new waste characterization, research, and regulatory information. However, recent budget guidance from DOE indicates that minimum funding will not be sufficient to accomplish original objectives to upgrade the liquid low-level waste (LLLW) system to comply with the Federal Facilities Agreement, provide long-term LLLW treatment capability, and minimize environmental, safety, and health risks. Options are presented that might allow the ORNL LLLW system to continue operations temporarily, but they would significantly reduce its capabilities to handle emergency situations, provide treatment for new waste streams, and accommodate waste from the Environmental Restoration Program and from decontamination and decommissioning of surplus facilities. These options are also likely to increase worker radiation exposure, risk of environmental insult, and generation of solid waste for on-site and off-site disposal/storage beyond existing facility capacities. The strategy will be fully developed after receipt of additional guidance. The proposed budget limitations are too severe to allow ORNL to meet regulatory requirements or continue operations long term.

Robinson, S.M.; Abraham, T.J.; DePaoli, S.M.; Walker, A.B.

1995-04-01T23:59:59.000Z

231

ALTERNATIVE THERMAL DESTRUCTION PROCESSES FOR HAZARDOUS WASTES  

E-Print Network (OSTI)

·Product Gas 400 2,000 11,300 Natural Gas 15,900 57,700 11,300 Most of these boilers are very small natural gas Distillate oil Natural gas Residual oil Distillate oil Natural gas Bituminous coal Bituminous coal Percent regulations. Candidate thermal processes include industrial processes such as boilers, process heaters, cement

Columbia University

232

TECHNOLOGY SUMMARY ADVANCING TANK WASTE RETREIVAL AND PROCESSING  

SciTech Connect

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.

SAMS TL

2010-07-07T23:59:59.000Z

233

TECHNOLOGY SUMMARY ADVANCING TANK WASTE RETRIEVAL AND PROCESSING  

SciTech Connect

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.

SAMS TL; MENDOZA RE

2010-08-11T23:59:59.000Z

234

Technology development program for Idaho Chemical Processing Plant spent fuel and waste management  

SciTech Connect

Acidic high-level radioactive waste (HLW) resulting from fuel reprocessing at the Idaho Chemical Processing Plant (ICPP) for the U.S. Department of Energy (DOE) has been solidified to a calcine since 1963 and stored in stainless steel bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage at the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, and describes the Spent Fuel and HLW Technology program in more detail.

Ermold, L.F.; Knecht, D.A.; Hogg, G.W.; Olson, A.L.

1993-08-01T23:59:59.000Z

235

Materials selection for process equipment in the Hanford waste vitrification plant  

Science Conference Proceedings (OSTI)

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.

Elmore, M R; Jensen, G A

1991-07-01T23:59:59.000Z

236

Process for blending coal with water immiscible liquid  

DOE Patents (OSTI)

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.

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

1982-10-26T23:59:59.000Z

237

High magnetic field processing of liquid crystalline polymers  

DOE Patents (OSTI)

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.

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

1998-11-24T23:59:59.000Z

238

High magnetic field processing of liquid crystalline polymers  

DOE Patents (OSTI)

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.

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

1998-01-01T23:59:59.000Z

239

Savannah River Site Salt Waste Processing Facility Technology Readiness Assessment Report  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Salt Waste Processing Facility Salt Waste Processing Facility Technology Readiness Assessment Report Kurt D. Gerdes Harry D. Harmon Herbert G. Sutter Major C. Thompson John R. Shultz Sahid C. Smith July 13, 2009 Prepared by the U.S. Department of Energy Washington, D.C. SRS Salt Waste Processing Facility Technology Readiness Assessment July 13, 2009 ii This page intentionally left blank SRS Salt Waste Processing Facility Technology Readiness Assessment July 13, 2009 iii SRS Salt Waste Processing Facility Technology Readiness Assessment July 13, 2009 iii Signatures SRS Salt Waste Processing Facility Technology Readiness Assessment July 13, 2009 iv This page intentionally left blank SRS Salt Waste Processing Facility

240

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

Science Conference Proceedings (OSTI)

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

JOHNSTON GA

2008-01-15T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Predictive LPV control of a liquid-gas separation process  

Science Conference Proceedings (OSTI)

The problem of controlling a liquid-gas separation process is approached by using LPV control techniques. An LPV model is derived from a nonlinear model of the process using differential inclusion techniques. Once an LPV model is available, an LPV controller ... Keywords: BMIs, LMIs, LPV controllers, LPV systems, Nonlinear systems, Predictive control

J. V. Salcedo; M. Martínez; C. Ramos; J. M. Herrero

2007-07-01T23:59:59.000Z

242

Savannah River Site - Salt Waste Processing Facility Independent Technical Review  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SALT WASTE PROCESSING FACILITY SALT WASTE PROCESSING FACILITY INDEPENDENT TECHNICAL REVIEW November 22, 2006 Conducted by: Harry Harmon, Team Lead Civil/Structural Sub Team Facility Safety Sub Team Engineering Sub Team Peter Lowry, Lead James Langsted, Lead George Krauter, Lead Robert Kennedy Chuck Negin Art Etchells Les Youd Jerry Evatt Oliver Block Loring Wyllie Richard Stark Tim Adams Tom Anderson Todd LaPointe Stephen Gosselin Carl Costantino Norman Moreau Patrick Corcoran John Christian Ken Cooper Kari McDaniel _____________________________ Harry D. Harmon ITR Team Leader SPD-SWPF-217 SPD-SWPF-217: Salt Waste Processing Facility Independent Technical Review 11/22/2006 ACKNOWLEDGEMENT The ITR Team wishes to thank Shari Clifford of Pacific Northwest National Laboratory for

243

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

Science Conference Proceedings (OSTI)

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.

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

1993-09-01T23:59:59.000Z

244

Audit of the radioactive liquid waste treatment facility operations at the Los Alamos National Laboratory  

SciTech Connect

Los Alamos National Laboratory (Los Alamos) generates radioactive and liquid wastes that must be treated before being discharged to the environment. Presently, the liquid wastes are treated in the Radioactive Liquid Waste Treatment Facility (Treatment Facility), which is over 30 years old and in need of repair or replacement. However, there are various ways to satisfy the treatment need. The objective of the audit was to determine whether Los Alamos cost effectively managed its Treatment Facility operations. The audit determined that Los Alamos` treatment costs were significantly higher when compared to similar costs incurred by the private sector. This situation occurred because Los Alamos did not perform a complete analysis of privatization or prepare a {open_quotes}make-or-buy{close_quotes} plan for its treatment operations, although a {open_quotes}make-or-buy{close_quotes} plan requirement was incorporated into the contract in 1996. As a result, Los Alamos may be spending $2.15 million more than necessary each year and could needlessly spend $10.75 million over the next five years to treat its radioactive liquid waste. In addition, Los Alamos has proposed to spend $13 million for a new treatment facility that may not be needed if privatization proves to be a cost effective alternative. We recommended that the Manager, Albuquerque Operations Office (Albuquerque), (1) require Los Alamos to prepare a {open_quotes}make-or-buy{close_quotes} plan for its radioactive liquid waste treatment operations, (2) review the plan for approval, and (3) direct Los Alamos to select the most cost effective method of operations while also considering other factors such as mission support, reliability, and long-term program needs. Albuquerque concurred with the recommendations.

1997-11-19T23:59:59.000Z

245

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

Science Conference Proceedings (OSTI)

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.

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

2009-01-11T23:59:59.000Z

246

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

Science Conference Proceedings (OSTI)

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.

Wilmarth, B; Sheryl Bush, S

2008-10-31T23:59:59.000Z

247

Process Design Concepts for Stabilization of High Level Waste Calcine  

Science Conference Proceedings (OSTI)

The current baseline assumption is that packaging ¡§as is¡¨ and direct disposal of high level waste (HLW) calcine in a Monitored Geologic Repository will be allowed. The fall back position is to develop a stabilized waste form for the HLW calcine, that will meet repository waste acceptance criteria currently in place, in case regulatory initiatives are unsuccessful. A decision between direct disposal or a stabilization alternative is anticipated by June 2006. The purposes of this Engineering Design File (EDF) are to provide a pre-conceptual design on three low temperature processes under development for stabilization of high level waste calcine (i.e., the grout, hydroceramic grout, and iron phosphate ceramic processes) and to support a down selection among the three candidates. The key assumptions for the pre-conceptual design assessment are that a) a waste treatment plant would operate over eight years for 200 days a year, b) a design processing rate of 3.67 m3/day or 4670 kg/day of HLW calcine would be needed, and c) the performance of waste form would remove the HLW calcine from the hazardous waste category, and d) the waste form loadings would range from about 21-25 wt% calcine. The conclusions of this EDF study are that: (a) To date, the grout formulation appears to be the best candidate stabilizer among the three being tested for HLW calcine and appears to be the easiest to mix, pour, and cure. (b) Only minor differences would exist between the process steps of the grout and hydroceramic grout stabilization processes. If temperature control of the mixer at about 80„aC is required, it would add a major level of complexity to the iron phosphate stabilization process. (c) It is too early in the development program to determine which stabilizer will produce the minimum amount of stabilized waste form for the entire HLW inventory, but the volume is assumed to be within the range of 12,250 to 14,470 m3. (d) The stacked vessel height of the hot process vessels in the hydroceramic grout process (i.e., 21 m) appears to be about the same as that estimated by the Direct Cementitious Waste Process in 1998, for which a conceptual design was developed. Some of the conceptual design efforts in the 1998 study may be applicable to the stabilizer processes addressed in this EDF. (e) The gamma radiation fields near the process vessels handling HLW calcine would vary from a range of about 300-350 R/hr at a distance of 2.5 cm from the side of the vessels to a range of about 50-170 R/hr at a distance of 100 cm from the side of the vessels. The calculations were made for combined calcine, which was defined as the total HLW calcine inventory uniformly mixed. (f) The gamma radiation fields near the stabilized waste in canisters would range from about 25-170 R/hr at 2.5 cm from the side of the canister and 5-35 R/hr at 100 cm from the side of the canister, depending on the which bin set was the source of calcine.

T. R. Thomas; A. K. Herbst

2005-06-01T23:59:59.000Z

248

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Salt Waste Processing Facility Independent Technical Review Harry Harmon January 9, 2007 2 U.S. Department of Energy Outline * SWPF Process Overview * Major Risks * Approach for Conducting Review * Discussion of Findings * Conclusions 3 U.S. Department of Energy Salt Waste Processing Facility 4 U.S. Department of Energy SWPF Process Overview Alpha Finishing Process CSSX Alpha Strike Process MST/ Sludge Cs Strip Effluent DSS 5 U.S. Department of Energy BOTTOM LINE The SWPF Project is ready to move into final design. 6 U.S. Department of Energy Major Risks * Final geotechnical data potentially could result in redesign of the PC-3 CPA base mat and structure. * Cost and schedule impacts arising from the change from ISO-9001 to NQA-1 quality assurance requirements. * The "de-inventory, flush, and then hands-on

249

Direction of CRT waste glass processing: Electronics recycling industry communication  

Science Conference Proceedings (OSTI)

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.

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

250

Technical Safety Requirements (TSR) for Waste Receiving & Processing (WRAP) facility  

SciTech Connect

These Technical Safety Requirements (TSRs) define the Administrative Controls required to ensure safe operation of the Waste Receiving and Processing Facility (WRAP). As will be shown in the report, Safety Limits, Limiting Control Settings, Limiting Conditions for Operation, and Surveillance Requirements are not required for safe operation of WRAP.

TOMASZEWSKI, T.A.

2001-07-10T23:59:59.000Z

251

Process for removal of sulfur oxides from waste gases  

Science Conference Proceedings (OSTI)

A process for removing sulfur oxides from waste gas is provided. The gas is contacted with a sorbent selected from sodium bicarbonate, trona and activated sodium carbonate and, utilizing an alkaline liquor containing borate ion so as to reduce flow rates and loss of alkalinity, the spent sorbent is regenerated with an alkaline earth metal oxide or hydroxide.

Lowell, P.S.; Phillips, J.L.

1983-05-24T23:59:59.000Z

252

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

SciTech Connect

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.

Van Hoesen, S.D.

2003-09-09T23:59:59.000Z

253

The largest radioactive waste glassification  

NLE Websites -- All DOE Office Websites (Extended Search)

largest radioactive waste glassification largest radioactive waste glassification plant in the nation, the Defense Waste Processing Facility (DWPF) converts the liquid nuclear waste currently stored at the Savannah River Site (SRS) into a solid glass form suitable for long-term storage and disposal. Scientists have long considered this glassification process, called "vitrification," as the preferred option for treating liquid nuclear waste. By immobilizing the radioactivity in glass, the DWPF reduces the risks associated with the continued storage of liquid nuclear waste at SRS and prepares the waste for final disposal in a federal repository. About 38 million gallons of liquid nuclear wastes are now stored in 49 underground carbon-steel tanks at SRS. This waste has about 300 million curies of radioactivity, of which the vast majority

254

Electrochemical processing of nitrate waste solutions  

SciTech Connect

The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F[sup [minus

Genders, D.; Weinberg, N.; Hartsough, D. (Electrosynthesis Co., Inc., Cheektowaga, NY (United States))

1992-10-07T23:59:59.000Z

255

Multi-step process for concentrating magnetic particles in waste sludges  

DOE Patents (OSTI)

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.

Watson, J.L.

1990-07-10T23:59:59.000Z

256

Waste Energy Analysis Recovery for a Typical Food Processing Plant  

E-Print Network (OSTI)

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. Additional analysis, measurements, and observations over a period of 12 months resulted in an evaluation of the losses as well as recommended methods for the effective recovery of the waste heat energy. The waste energy recovery results in significant savings in energy costs as well as a reduction in the consumption of scarce fuel. The research was supported by the Louisiana Department of Natural Resources, College of Engineering, Louisiana State University, and the Joan of Arc Company, St. Francisville, Louisiana.

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

1980-01-01T23:59:59.000Z

257

Summary - SRS Salt Waste Processing Facility  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SRS Co SRS Co DOE S Proces concen actinid in a se remov adjustm sorben sorben solutio passed separa stream extract sufficie separa (with S vitrifica (DWP Sr/acti federa assure and ha Critica The te (CTE) descrip Readin The Ele Site: S roject: S F Report Date: J ited States Why DOE omposite High Lev Savannah Rive ssing Facility (S ntrate targeted des) from High eries of unit ope ved by contactin ment) with a m nt in a batch m nt (containing S on by cross flow d to a solvent e ated to an aque m. The bulk so tion process, w ently low levels ated high activi Sr and actinide ation in the Def F). Provisions inides adsorpti al project direct e that the plann ave been matu al Decision-3 ap What th eam identified e of the SWPF w ption. All CTE ness Level of 6 To view the full T http://www.em.doe. objective of a Tech ements (CTEs), usin

258

Zero-Release Mixed Waste Process Facility Design and Testing  

SciTech Connect

A zero-release offgas cleaning system for mixed-waste thermal treatment processes has been evaluated through experimental scoping tests and process modeling. The principles can possibly be adapted to a fluidized-bed calcination or stream reforming process, a waste melter, a rotarykiln process, and possibly other waste treatment thermal processes. The basic concept of a zero-release offgas cleaning system is to recycle the bulk of the offgas stream to the thermal treatment process. A slip stream is taken off the offgas recycle to separate and purge benign constituents that may build up in the gas, such as water vapor, argon, nitrogen, and CO2. Contaminants are separated from the slip stream and returned to the thermal unit for eventual destruction or incorporation into the waste immobilization media. In the current study, a standard packed-bed scrubber, followed by gas separation membranes, is proposed for removal of contaminants from the offgas recycle slipstream. The scrub solution is continuously regenerated by cooling and precipitating sulfate, nitrate, and other salts that reach a solubility limit in the scrub solution. Mercury is also separated by the scrubber. A miscible chemical oxidizing agent was shown to effectively oxidize mercury and also NO, thus increasing their removal efficiency. The current study indicates that the proposed process is a viable option for reducing offgas emissions. Consideration of the proposed closed-system offgas cleaning loop is warranted when emissions limits are stringent, or when a reduction in the total gas emissions volume is desired. Although the current closed-loop appears to be technically feasible, economical considerations must be also be evaluated on a case-by-case basis.

Richard D. Boardman; John A. Deldebbio; Robert J. Kirkham; Martin K. Clemens; Robert Geosits; Ping Wan

2004-02-01T23:59:59.000Z

259

Processing of solid mixed waste containing radioactive and hazardous materials  

DOE Patents (OSTI)

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.

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

1998-05-12T23:59:59.000Z

260

Processing of solid mixed waste containing radioactive and hazardous materials  

DOE Patents (OSTI)

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.

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

1998-05-12T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
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261

Pyrochemical treatment of Idaho Chemical Processing Plant high-level waste calcine  

SciTech Connect

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1951 to recover uranium, krypton-85, and isolated fission products for interim treatment and immobilization. The acidic radioactive high-level liquid waste (HLLW) is routinely stored in stainless steel tanks and then, since 1963, calcined to form a dry granular solid. The resulting high-level waste (HLW) calcine is stored in seismically hardened stainless steel bins that are housed in underground concrete vaults. A research and development program has been established to determine the feasibility of treating ICPP HLW calcine using pyrochemical technology.This technology is described.

Todd, T.A.; DelDebbio, J.A.; Nelson, L.O.; Sharpsten, M.R.

1993-06-01T23:59:59.000Z

262

Drop Dynamics and Speciation in Isolation of Metals from Liquid Wastes by Reactive Scavenging  

SciTech Connect

Computational and experimental studies of the motion and dynamics of liquid drops in gas flows were conducted with relevance to reactive scavenging of metals from atomized liquid waste. Navier-Stoke's computations of deformable drops revealed a range of conditions from which prolate drops are expected, and showed how frajectiones of deformable drops undergoing deceleration can be computed. Experimental work focused on development of emission fluorescence, and scattering diagnostics. The instrument developed was used to image drop shapes, soot, and nonaxisymmetric departures from steady flow in a 22kw combustor

Arne J. Pearlstein; Alexander Scheeline

2002-08-30T23:59:59.000Z

263

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

DOE Green Energy (OSTI)

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.

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

2011-01-12T23:59:59.000Z

264

Development of ultrafiltration and inorganic adsorbents for reducing volumes of low-level and intermediate-level liquid waste. Quarterly report, October, November, December 1976  

SciTech Connect

The objective of this program is to develop and demonstrate separation methods for removing radionuclides from liquid process waste streams. As part of this program, Mound Laboratory will develop lower cost alternatives for use i n1980 fuel reprocessing and waste solidification plants, evaluate the processes within the nuclear fuel cycle which contribute to low-level and intermediate-level waste, and determine the feasibility of ultrafiltration, reverse osmosis, inorganic adsorbents and other separation concepts as additions to process design to reduce the generation of this type of waste. In the initial phase of this program, membrane equipment will be obtained from a commercial membrane manufacturer. After the pilot plant is installed, it will be checked out on cold feed in order to obtain initial flux and rejection data for comparison to data obtained later in the program. After completion of the cold tests, the membrane pilot plant will be run on a combined contaminated feed emanating from showers, laboratory drains, janitorial sinks and decontamination in processing areas, as well as a laundry waste stream containing alpha-contaminated wastes. This combined waste stream contains only alpha contamination (uranium and plutonium). However, as part of this program, gamma activity will be added to the waste stream. These wastes will be representative of those streams found at fuel reprocessing plants, as well as various ERDA processing facilities such as Mound, LASL, Hanford, and Rocky Flats. For the second part of the program, laboratory tests will be run on various adsorbents to evaluate their capacities for removing radionuclides. As part of this program, a technique for screening adsorbents developed at Mound Laboratory will be utilized.

1976-12-31T23:59:59.000Z

265

Demonstration of the TRUEX process for the treatment of actual high activity tank waste at the INEEL using centrifugal contactors  

SciTech Connect

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering and Environmental Laboratory (INEEL), formerly reprocessed spent nuclear fuel to recover fissionable uranium. The radioactive raffinates from the solvent extraction uranium recovery processes were converted to granular solids (calcine) in a high temperature fluidized bed. A secondary liquid waste stream was generated during the course of reprocessing, primarily from equipment decontamination between campaigns and solvent wash activities. This acidic tank waste cannot be directly calcined due to the high sodium content and has historically been blended with reprocessing raffinates or non-radioactive aluminum nitrate prior to calcination. Fuel reprocessing activities are no longer being performed at the ICPP, thereby eliminating the option of waste blending to deplete the waste inventory. Currently, approximately 5.7 million liters of high-activity waste are temporarily stored at the ICPP in large underground stainless-steel tanks. The United States Environmental Protection Agency and the Idaho Department of Health and Welfare filed a Notice of Noncompliance in 1992 contending some of the underground waste storage tanks do not meet secondary containment. As part of a 1995 agreement between the State of Idaho, the Department of Energy, and the Department of Navy, the waste must be removed from the tanks by 2012. Treatment of the tank waste inventories by partitioning the radionuclides and immobilizing the resulting high-activity and low-activity waste streams is currently under evaluation. A recent peer review identified the most promising radionuclide separation technologies for evaluation. The Transuranic Extraction-(TRUEX) process was identified as a primary candidate for separation of the actinides from ICPP tank waste.

Law, J.D.; Brewer, K.N.; Todd, T.A.; Olson, L.G.

1997-10-01T23:59:59.000Z

266

Current Practices: Solid Waste Management from Zero Liquid Discharge (ZLD) Wastewater Treatment  

Science Conference Proceedings (OSTI)

A study was conducted to identify current practices used by power plants to manage their solid waste residuals from zero liquid discharge (ZLD) operations treating flue gas desulfurization (FGD) wastewater. Because there are such few FGD ZLD systems in operation not only in the United States but also worldwide, the study scope was expanded to include non-FGD ZLD operations, as well. Only two of the seven facilities interviewed in this study operate ZLDs on FGD water; therefore, much of the current ...

2012-12-31T23:59:59.000Z

267

Process modeling of hydrogen production from municipal solid waste  

DOE Green Energy (OSTI)

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.

Thorsness, C.B.

1995-01-01T23:59:59.000Z

268

Lab Ahead of Schedule Processing Waste in Large Boxes | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Lab Ahead of Schedule Processing Waste in Large Boxes Lab Ahead of Schedule Processing Waste in Large Boxes Lab Ahead of Schedule Processing Waste in Large Boxes March 30, 2012 - 12:00pm Addthis A framework agreement between DOE and the State of New Mexico calls for the Lab’s TRU Waste Program to ship 3,706 cubic meters of combustible or dispersible transuranic waste to WIPP for permanent disposal by June 30, 2014. A framework agreement between DOE and the State of New Mexico calls for the Lab's TRU Waste Program to ship 3,706 cubic meters of combustible or dispersible transuranic waste to WIPP for permanent disposal by June 30, 2014. Processing waste in large boxes is ahead of schedule due to worker skill, efficient processing and good planning. Processing waste in large boxes is ahead of schedule due to worker skill,

269

THE USE OF POLYMERS IN RADIOACTIVE WASTE PROCESSING SYSTEMS  

SciTech Connect

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.

Skidmore, E.; Fondeur, F.

2013-04-15T23:59:59.000Z

270

Idaho Site Taps Old World Process to Treat Nuclear Waste | Department...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Idaho Site Taps Old World Process to Treat Nuclear Waste Idaho Site Taps Old World Process to Treat Nuclear Waste September 9, 2013 - 12:00pm Addthis The Idaho site's sodium...

271

Technical evaluation of the waste-to-oil process development facility at Albany, Oregon  

DOE Green Energy (OSTI)

The broad objective of ERDA's solar energy program at Albany, Oregon, is to develop biomass-to-synfuel technology in the Albany process development facility, which is now nearing completion. In the study reported here, the process development plant design was reevaluated, and a number of modifications and additions are recommended to facilitate and accelerate development of biomass conversion processes. Sketches of the recommended modifications and estimates of costs and installation time schedules have been provided. It has been found expedient to implement some of these modifications before construction is completed. Biomass-to-synfuel processes under development or under consideration elsewhere have been reviewed, and some have been identified that are appropriate for further development at Albany. Potential environmental impacts associated with the operation of the Albany, Oregon, facility were reviewed to identify the magnitude of the impacts and the effects of any resultant operational constraints. Two discrete environmental impact categories have been identified with respect to process development operation. These are (1) production, storage, and disposal of product oil and residual solid, liquid, and gaseous waste; and (2) disturbances to the local community. An assessment has been made of unit process waste discharges and mitigation procedures, environmental setting and community considerations, possible operational constraints, and monitoring programs.

Houle, E.H.; Ciriello, S.F.; Ergun, S.; Basuino, D.J.

1976-10-01T23:59:59.000Z

272

Out-Of-Drum Grout Mixer Testing With Simulated Liquid Effluents Originating From Sodium-Bearing Waste at the Idaho Nuclear Technology and Engineering Center  

SciTech Connect

The Idaho National Engineering and Environmental Laboratory (INEEL) is considering several optional processes for disposal of liquid sodium-bearing waste. During fiscal year 2003, alternatives were evaluated for grout formulation development and associated mixing for the Sodium-Bearing Waste cesium ion exchange process. The neutralization agents calcium or sodium hydroxide and the solidification agents Portland cement, with or without blast furnace slag were evaluated. A desired uniform formulation was pursued to develop a grout waste form without any bleed liquid and solidify within a reasonable period of about twenty-eight days. This testing evaluates the out-of-drum alternative of mixing the effluent with solidification agents prior to being poured into drums versus the in-drum alternative of mixing them all together after being poured into the drums. Experimental results indicate that sodium-bearing waste can be immobilized in grout using the Autocon continuous mixer within the range of 66 to 72 weight percent. Furthermore, a loading of 30 weight percent NWCF scrubber simulant also produced an acceptable grout waste form.

B. A. Scholes; A. K. Herbst; S. V. Raman; S. H. Hinckley

2003-09-01T23:59:59.000Z

273

Lessons learned in TRU waste process improvement at LANL  

SciTech Connect

Typical papers that discuss lessons learned or quality improvement focus on the challenge for a production facility reaching six sigma (3.4 Defects Per Million Opportunities) from five sigma. This paper discusses lessons learned when the Los Alamos National Laboratory's (LANL) transuranic (TRU) waste management project was challenged to establish a production system to meet the customer's expectations. The target for FY 2003 was set as two shipments of TRU waste per week leaving the site. The average for the four previous years (FY99-02) was about one shipment every two months. LANL recognized that, despite its success in 1999 as the first site to ship TRU waste to open the Waste Isolation Pilot Plant (WIPP), significant changes to the way business was being done were required to move to a production mode. Process improvements began in earnest in April 2002. This paper discusses several of the initiatives LANL took to achieve forty-five shipments in FY03. The paper is organized by topic into five major areas that LANL worked to get the job done.

Del Signore, J. C.; Huchton, J. (Judith); Martin, B. (Beverly); Lindahl, P. (Peter); Miller, S. (Scott); Hartwell, W. B. (Ware B.)

2004-01-01T23:59:59.000Z

274

EVALUATION AND SELECTION OF 99TC GETTERS FOR SEQUESTRATION OF LIQUID SECONDARY WASTE RESULTING FROM VITRIFICATION OF RADIOACTIVE WASTE FROM HANFORD  

Science Conference Proceedings (OSTI)

Getters are most commonly inorganic materials that selectively adsorb radionuclide and metallic contaminants. Typically, these materials have been deployed in two different modes to immobilize and retard contaminant release from monolithic waste forms. One mode is to first use getters to selectively scavenge the radionuclide of interest from a liquid waste stream, and then incorporate the radionuclide-loaded getters in cementitious or other monolithic waste forms. The other mode consists of mixing getters and liquid waste together during formulation of monolithic waste forms. Desirable characteristics for a getter material include, (1) specific adsorption of radionuclide of interest and very high selectivity toward radionuclides of concern in concentrations that would be several orders of magnitude less than the concentrations of competing anions and cations, (2) adsorption capacity that should be sufficient for the mass and volume of the material that will be deployed to be within practicable limits, (3) long-term adsorption and retention of radionuclide, (4) sufficient physical and chemical stability that its radionuclide retention performance will not degrade significantly during the designed life span of the waste form, (5) chemical stability under the range of Eh, pH, and solution conditions that exist in the waste form environment, and (6) should not adversely affect chemical and physical integrity of waste forms. We conducted a literature review to identify getters that are suitable for effectively sequestering 99Tc in monolithic waste forms that are being evaluated for stabilizing secondary liquid waste streams resulting from treatment and vitrification of radioactive tank wastes at Hanford. As a result of this review, we identified a set of getters that warrant further evaluation for this specific application.

Mattigod, Shas V.; Westsik, Joseph H.

2011-03-31T23:59:59.000Z

275

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

SciTech Connect

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.

Not Available

1994-11-01T23:59:59.000Z

276

Solidification of Simulated Liquid Effluents Originating From Sodium-Bearing Waste at the Idaho Nuclear Technology and Engineering Center, FY-03 Report  

SciTech Connect

In this report, the mechanism and methods of fixation of acidic waste effluents in grout form are explored. From the variations in the pH as a function of total solids addition to acidic waste effluent solutions, the stages of gellation, liquefaction, slurry formation and grout development are quantitatively revealed. Experimental results indicate the completion of these reaction steps to be significant for elimination of bleed liquid and for setting of the grout to a dimensionally stable and hardened solid within a reasonable period of about twenty eight days that is often observed in the cement and concrete industry. The reactions also suggest increases in the waste loading in the direction of decreasing acid molarity. Consequently, 1.0 molar SBW-180 waste is contained in higher quantity than the 2.8 molar SBW-189, given the same grout formulation for both effluents. The variations in the formulations involving components of slag, cement, waste and neutralizing agent are represented in the form of a ternary formulation map. The map in turn graphically reveals the relations among the various formulations and grout properties, and is useful in predicting the potential directions of waste loading in grouts with suitable properties such as slurry viscosity, Vicat hardness, and mechanical strength. A uniform formulation for the fixation of both SBW-180 and SBW-189 has emerged from the development of the formulation map. The boundaries for the processing regime on this map are 100 wt% cement to 50 wt% cement / 50 wt% slag, with waste loadings ranging from 55 wt% to 68 wt%. Within these compositional bounds all the three waste streams SBW-180, SBW-189 and Scrub solution are amenable to solidification. A large cost advantage is envisaged to stem from savings in labor, processing time, and processing methodology by adopting a uniform formulation concept for fixation of compositionally diverse waste streams. The experimental efforts contained in this report constitute the first attempt at developing a uniform methodology.

S. V. Raman; A. K. Herbst; B. A. Scholes; S. H. Hinckley; R. D. Colby

2003-09-01T23:59:59.000Z

277

Process for hydrocracking carbonaceous material in liquid carrier  

DOE Patents (OSTI)

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.

Duncan, Dennis A. (Downers Grove, IL)

1980-01-01T23:59:59.000Z

278

New process effectively recovers oil from refinery waste streams  

Science Conference Proceedings (OSTI)

A new process uses chemically assisted, thermal flashing to break difficult emulsions and recover oil for reprocessing. The process is best suited for refinery waste management and slop oil systems, where it can process streams with high oil content to recover high-quality oil. Recent testing of a full-scale, commercial prototype unit on slop oil emulsions at a major Gulf Coast refinery resulted in: 97.9% recovery of oil with 99.3--99.6% purity; 99.5% recovery of water with 99+% purity; and a centrifuge cake containing 49-60% solids, 23--30 oil, and 17--22% water. The paper discusses background of the process, then gives a process description as well as results of field studies and cost.

Rhodes, A.

1994-08-15T23:59:59.000Z

279

Secondary Waste Forms and Technetium Management  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

November 18, 2010 What are Secondary Wastes? Process condensates and scrubber andor off-gas treatment liquids from the pretreatment and ILAW melter facilities at the Hanford WTP....

280

GRR/Section 18-CO-b - Hazardous Waste Permit Process | Open Energy  

Open Energy Info (EERE)

GRR/Section 18-CO-b - Hazardous Waste Permit Process GRR/Section 18-CO-b - Hazardous Waste Permit Process < GRR Jump to: navigation, search GRR-logo.png GEOTHERMAL REGULATORY ROADMAP Roadmap Home Roadmap Help List of Sections Section 18-CO-b - Hazardous Waste Permit Process 18COBHazardousWastePermitProcess.pdf Click to View Fullscreen Contact Agencies Colorado Department of Public Health and Environment Regulations & Policies Colorado Hazardous Waste Regulations Part 260 Triggers None specified Click "Edit With Form" above to add content 18COBHazardousWastePermitProcess.pdf 18COBHazardousWastePermitProcess.pdf Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Flowchart Narrative Hazardous waste is a regulated substance and facilities that treat, store

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281

GRR/Section 18-ID-b - Hazardous Waste Permit Process | Open Energy  

Open Energy Info (EERE)

GRR/Section 18-ID-b - Hazardous Waste Permit Process GRR/Section 18-ID-b - Hazardous Waste Permit Process < GRR Jump to: navigation, search GRR-logo.png GEOTHERMAL REGULATORY ROADMAP Roadmap Home Roadmap Help List of Sections Section 18-ID-b - Hazardous Waste Permit Process 18IDBHazardousWastePermitProcess.pdf Click to View Fullscreen Contact Agencies Idaho Department of Environmental Quality Regulations & Policies Idaho Hazardous Waste Management Act IDAPA 58.01.05 Rules and Standards for Hazardous Waste 40 CFR 124.31 Pre-application public meeting and notice 40 CRF 124.10 Public notice of permit actions and public comment period 40 CFR 124.12 Public hearings 40 CFR 270.13 Contents of Part A of the permit application Triggers None specified Click "Edit With Form" above to add content 18IDBHazardousWastePermitProcess.pdf 18IDBHazardousWastePermitProcess.pdf

282

Indirect thermal liquefaction process for producing liquid fuels from biomass  

DOE Green Energy (OSTI)

A progress report on an indirect liquefaction process to convert biomass type materials to quality liquid hydrocarbon fuels by gasification followed by catalytic liquid fuels synthesis has been presented. A wide variety of feedstocks can be processed through the gasification system to a gas with a heating value of 500 + Btu/SCF. Some feedstocks are more attractive than others with regard to producing a high olefin content. This appears to be related to hydrocarbon content of the material. The H/sub 2//CO ratio can be manipulated over a wide range in the gasification system with steam addition. Some feedstocks require the aid of a water-gas shift catalyst while others appear to exhibit an auto-catalytic effect to achieve the conversion. H/sub 2/S content (beyond the gasification system wet scrubber) is negligible for the feedstocks surveyed. The water gas shift reaction appears to be enhanced with an increase in pyrolysis reactor temperature over the range of 1300 to 1700/sup 0/F. Reactor temperature in the Fischer-Tropsch step is a significant factor with regard to manipulating product composition analysis. The optimum temperature however will probably correspond to maximum conversion to liquid hydrocarbons in the C/sub 5/ - C/sub 17/ range. Continuing research includes integrated system performance assessment, alternative feedstock characterization (through gasification) and factor studies for gasification (e.g., catalyst usage, alternate heat transfer media, steam usage, recycle effects, residence time study) and liquefaction (e.g., improved catalysts, catalyst activity characterization).

Kuester, J.L.

1980-01-01T23:59:59.000Z

283

NGNP Process Heat Utilization: Liquid Metal Phase Change Heat Exchanger  

DOE Green Energy (OSTI)

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.

Piyush Sabharwall; Mike Patterson; Vivek Utgikar; Fred Gunnerson

2008-09-01T23:59:59.000Z

284

Defense Waste Processing Facility -- Radioactive operations -- Part 3 -- Remote operations  

SciTech Connect

The Savannah River Site`s Defense Waste Processing Facility (DWPF) near Aiken, South Carolina is the nation`s first and world`s largest vitrification facility. Following a ten year construction period and nearly three years of non-radioactive testing, the DWPF began radioactive operations in March 1996. Radioactive glass is poured from the joule heated melter into the stainless steel canisters. The canisters are then temporarily sealed, decontaminated, resistance welded for final closure, and transported to an interim storage facility. All of these operations are conducted remotely with equipment specially designed for these processes. This paper reviews canister processing during the first nine months of radioactive operations at DWPF. The fundamental design consideration for DWPF remote canister processing and handling equipment are discussed as well as interim canister storage.

Barnes, W.M.; Kerley, W.D.; Hughes, P.D.

1997-06-01T23:59:59.000Z

285

CROWTM PROCESS APPLICATION FOR SITES CONTAMINATED WITH LIGHT NON-AQUEOUS PHASE LIQUIDS AND CHLORINATED HYDROCARBONS  

DOE Green Energy (OSTI)

Western Research Institute (WRI) has successfully applied the CROWTM (Contained Recovery of Oily Wastes) process at two former manufactured gas plants (MGPs), and a large wood treatment site. The three CROW process applications have all occurred at sites contaminated with coal tars or fuel oil and pentachlorophenol (PCP) mixtures, which are generally denser than water and are classified as dense non-aqueous phase liquids (DNAPLs). While these types of sites are abundant, there are also many sites contaminated with gasoline, diesel fuel, or fuel oil, which are lighter than water and lie on top of an aquifer. A third site type occurs where chlorinated hydrocarbons have contaminated the aquifer. Unlike the DNAPLs found at MGP and wood treatment sites, chlorinated hydrocarbons are approximately one and a half times more dense than water and have fairly low viscosities. These contaminants tend to accumulate very rapidly at the bottom of an aquifer. Trichloroethylene (TCE) and perchloroethylene, or tetrachloroethylene (PCE), are the major industrial chlorinated solvents that have been found contaminating soils and aquifers. The objective of this program was to demonstrate the effectiveness of applying the CROW process to sites contaminated with light non-aqueous phase liquids (LNAPLs) and chlorinated hydrocarbons. Individual objectives were to determine a range of operating conditions necessary to optimize LNAPL and chlorinated hydrocarbon recovery, to conduct numerical simulations to match the laboratory experiments and determine field-scale recoveries, and determine if chemical addition will increase the process efficiency for LNAPLs. The testing consisted of twelve TCE tests; eight tests with PCE, diesel, and wood treatment waste; and four tests with a fuel oil-diesel blend. Testing was conducted with both vertical and horizontal orientations and with ambient to 211 F (99 C) water or steam. Residual saturations for the horizontal tests ranged from 23.6% PV to 0.3% PV. Also conducted was screening of 13 chemicals to determine their relative effectiveness and the selection of three chemicals for further testing.

L.A. Johnson, Jr.

2003-06-30T23:59:59.000Z

286

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

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.

Aaron, G.; Wilmarth, B.

2011-09-19T23:59:59.000Z

287

Improved Process Used to Treat Aqueous Mixed Waste Results in Cost Savings and Improved Worker Safety  

Science Conference Proceedings (OSTI)

This paper describes an improved process implemented at Argonne National Laboratory (ANL) to treat aqueous mixed waste. This waste is comprised of radioactively-contaminated corrosive liquids with heavy metals. The Aqueous Mixed Waste Treatment System (AMWTS) system components include a reaction tank and a post-treatment holding tank with ancillary piping and pumps; and a control panel with pumping/mixing controls; tank level, temperature and pH/Oxidation Reduction Potential (ORP) indicators. The process includes a neutralization step to remove the corrosive characteristic, a chromium reduction step to reduce hexavalent chromium to trivalent chromium, and a precipitation step to convert the toxic metals into an insoluble form. Once the toxic metals have precipitated, the resultant sludge is amenable to stabilization and can be reclassified as a low-level waste if the quantity of leachable toxic metals, as determined by the TCLP, is below Universal Treatment Standards (UTS). To date, six batches in eight have passed the UTS. The AMWTS is RCRA permitted and allows for the compliant treatment of mixed waste prior to final disposal at a Department of Energy (DOE) or commercial radioactive waste disposal facility. Mixed wastes eligible for treatment include corrosive liquids (pH 12.5) containing EPA-regulated toxic metals (As, Ba, Pb, Cd, Cr, Ag, Se, Hg) at concentrations greater than the RCRA Toxicity Characteristic Leaching Procedure (TCLP) limit. The system has also been used to treat corrosive wastes with small quantities of fissionable materials. The AMWTS is a significant engineered solution with many improvements over the more labor intensive on-site treatment method being performed within a ventilation hood used previously. The previously used treatment system allowed for batch sizes of only 15-20 gallons whereas the new AMWTS allows for the treatment of batches up to 75 gallons; thereby reducing batch labor and supply costs by 40-60% and reducing analytical testing costs by 50-75%. Reduced treatment time also reduces worker radiation exposure to As Low As Reasonably Achievable (ALARA) levels. Additionally, the treatment system components used previously were adapted to be used with the new AMWTS. This allowed for less dependence on personnel protective equipment (PPE) than the prior system by separating the waste handling/bulking steps of the process from the treatment steps. The AMWTS also improved worker safety by incorporating more automated engineering controls such as system logic controls; personnel safety and equipment protection interlocks, off normal condition indicators/alarms, and system emergency stop controls. In a time of ever-decreasing budgets, it makes sense to rethink the use of existing treatment systems. Utilizing, and possibly retooling, equipment and infrastructure may allow for reduced treatment costs and increase worker safety. (authors)

Hodge, D.S.; Preuss, D.E.; Belcher, K.J.; Rock, C.M.; Bray, W.S.; Herman, J.P. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2006-07-01T23:59:59.000Z

288

NATURE OF RADIOACTIVE WASTES  

SciTech Connect

The integrated processes of nuclear industry are considered to define the nature of wastes. Processes for recovery and preparation of U and Th fuels produce wastes containing concentrated radioactive materials which present problems of confinement and dispersal. Fundamentals of waste treatment are considered from the standpoint of processes in which radioactive materials become a factor such as naturally occurring feed materials, fission products, and elements produced by parasitic neutron capture. In addition, the origin of concentrated fission product wastes is examined, as well as characteristics of present wastes and the level of fission products in wastes. Also, comments are included on high-level wastes from processes other than solvent extraction, active gaseous wastes, and low- to intermediate-level liquid wastes. (J.R.D.)

Culler, F.L. Jr.

1959-01-26T23:59:59.000Z

289

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

SciTech Connect

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.

Lambert, D.P.

2000-03-22T23:59:59.000Z

290

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

Science Conference Proceedings (OSTI)

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.

NONE

1997-01-01T23:59:59.000Z

291

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

SciTech Connect

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.

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

292

GRR/Section 18-UT-b - Hazardous Waste Permit Process | Open Energy  

Open Energy Info (EERE)

UT-b - Hazardous Waste Permit Process UT-b - Hazardous Waste Permit Process < GRR Jump to: navigation, search GRR-logo.png GEOTHERMAL REGULATORY ROADMAP Roadmap Home Roadmap Help List of Sections Section 18-UT-b - Hazardous Waste Permit Process 18UTBHazardousWastePermitProcess (1).pdf Click to View Fullscreen Contact Agencies Utah Department of Environmental Quality Regulations & Policies Hazardous Waste Rules R315-1 et seq Triggers None specified Click "Edit With Form" above to add content 18UTBHazardousWastePermitProcess (1).pdf Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Flowchart Narrative A hazardous waste is specifically listed by the Utah Solid and Hazardous Waste Rules or exhibits a characteristic such as ignitability, corrosivity,

293

Summary - Salt Waste Processing Facility Design at the Savannah River Site  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Salt Waste Processing Facility Salt Waste Processing Facility ETR Report Date: November 2006 ETR-4 United States Department of Energy Office of Environmental Management (DOE-EM) External Technical Review of the Salt Waste Processing Facility Design at the Savannah River Site (SRS) Why DOE-EM Did This Review The Salt Waste Processing Facility (SWPF) is intended to remove and concentrate the radioactive strontium (Sr), actinides, and cesium (Cs) from the bulk salt waste solutions in the SRS high-level waste tanks. The sludge and strip effluent from the SWPF that contain concentrated Sr, actinide, and Cs wastes will be sent to the SRS Defense Waste Processing Facility (DWPF), where they will be vitrified. The decontaminated salt solution (DSS) that is left after removal of the highly

294

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

SciTech Connect

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.

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

2006-04-17T23:59:59.000Z

295

Melt processing of radioactive waste: A technical overview  

Science Conference Proceedings (OSTI)

Nuclear operations have resulted in the accumulation of large quantities of contaminated metallic waste which are stored at various DOE, DOD, and commercial sites under the control of DOE and the Nuclear Regulatory Commission (NRC). This waste will accumulate at an increasing rate as commercial nuclear reactors built in the 1950s reach the end of their projected lives, as existing nuclear powered ships become obsolete or unneeded, and as various weapons plants and fuel processing facilities, such as the gaseous diffusion plants, are dismantled, repaired, or modernized. For example, recent estimates of available Radioactive Scrap Metal (RSM) in the DOE Nuclear Weapons Complex have suggested that as much as 700,000 tons of contaminated 304L stainless steel exist in the gaseous diffusion plants alone. Other high-value metals available in the DOE complex include copper, nickel, and zirconium. Melt processing for the decontamination of radioactive scrap metal has been the subject of much research. A major driving force for this research has been the possibility of reapplication of RSM, which is often very high-grade material containing large quantities of strategic elements. To date, several different single and multi-step melting processes have been proposed and evaluated for use as decontamination or recycling strategies. Each process offers a unique combination of strengths and weaknesses, and ultimately, no single melt processing scheme is optimum for all applications since processes must be evaluated based on the characteristics of the input feed stream and the desired output. This paper describes various melt decontamination processes and briefly reviews their application in developmental studies, full scale technical demonstrations, and industrial operations.

Schlienger, M.E.; Buckentin, J.M.; Damkroger, B.K.

1997-04-01T23:59:59.000Z

296

RPP-PLAN-47325 Revision 0 Radioactive Waste Determination Process Plan for Waste Management Area C Tank  

E-Print Network (OSTI)

This plan describes the radioactive waste determination process that the U.S. Department of Energy (DOE) will use for Hanford Site Waste Management Area C (WMA C) tank waste residuals subject to DOE authority under DOE Order 435.1, Radioactive Waste Management. Preparation of this plan is a required component of actions the DOE-Office of River Protection (ORP) must take to fulfill proposed Hanford Federal Facility Agreement and Consent Order Milestone M-045-80. Waste Management Area C is comprised of various single-shell tanks, encased and direct-buried pipes, diversion boxes, pump pits, and unplanned release sites (sites contaminated as a result of spills of tank waste to the environment). Since operations began in the late 1940s, the tanks in WMA C have continuously stored waste managed as high-level waste (HLW) that was derived from defense-related nuclear research, development, and weapons production activities. Planning for the final closure of WMA C is underway. This radioactive waste determination process plan assumes that tank closure will follow retrieval of as much tank waste as technically and economically practical. It is also assumed for the purposes of this plan that after completion

Waste Residuals; J. R. Robertson

2010-01-01T23:59:59.000Z

297

Feed Composition for Sodium-Bearing Waste Treatment Process  

SciTech Connect

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.

Barnes, C.M.

2000-10-30T23:59:59.000Z

298

Process of producing liquid hydrocarbon fuels from biomass  

DOE Patents (OSTI)

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+.

Kuester, James L. (Scottsdale, AZ)

1987-07-07T23:59:59.000Z

299

Process of producing liquid hydrocarbon fuels from biomass  

DOE Patents (OSTI)

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.

Kuester, J.L.

1987-07-07T23:59:59.000Z

300

Environmental sampling program for a solar evaporation pond for liquid radioactive wastes  

Science Conference Proceedings (OSTI)

Los Alamos Scientific Laboratory (LASL) is evaluating solar evaporation as a method for disposal of liquid radioactive wastes. This report describes a sampling program designed to monitor possible escape of radioactivity to the environment from a solar evaporation pond prototype constructed at LASL. Background radioactivity levels at the pond site were determined from soil and vegetation analyses before construction. When the pond is operative, the sampling program will qualitatively and quantitatively detect the transport of radioactivity to the soil, air, and vegetation in the vicinity. Possible correlation of meteorological data with sampling results is being investigated and measures to control export of radioactivity by biological vectors are being assessed.

Romero, R.; Gunderson, T.C.; Talley, A.D.

1980-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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301

New methods for determination of interstitial liquid levels in Hanford waste tanks  

Science Conference Proceedings (OSTI)

The key to the leak detection program for many tanks at Hanford is the method used to evaluate the apparent interstitial liquid interface (ILL) within the pore space of the solid waste medium (either crystalline or sludge). Three new approaches were introduced in the summer of 1993 (count rate, derivative, and sigmoid), all of which significantly improved the accuracy and repeatability of interstitial liquid level values from neutron survey data. This paper summarizes the three new methods and details a case study in which, as a direct result of this improved analysis, a tank that had been declared an ``assumed leaker`` was reclassified as ``sound`` for the first time in Hanford`s 50 year history.

Barnes, D.A.; Raymond, R.E. [Westinghouse Hanford Co., Richland, WA (United States); Whitney, P.D. [Pacific Northwest Lab., Richland, WA (United States)

1995-01-01T23:59:59.000Z

302

DETERMINATION OF LIQUID FILM THICKNESS FOLLOWING DRAINING OF CONTACTORS, VESSELS, AND PIPES IN THE MCU PROCESS  

Science Conference Proceedings (OSTI)

The Department of Energy (DOE) identified the caustic side solvent extraction (CSSX) process as the preferred technology to remove cesium from radioactive waste solutions at the Savannah River Site (SRS). As a result, Washington Savannah River Company (WSRC) began designing and building a Modular CSSX Unit (MCU) in the SRS tank farm to process liquid waste for an interim period until the Salt Waste Processing Facility (SWPF) begins operations. Both the solvent and the strip effluent streams could contain high concentrations of cesium which must be removed from the contactors, process tanks, and piping prior to performing contactor maintenance. When these vessels are drained, thin films or drops will remain on the equipment walls. Following draining, the vessels will be flushed with water and drained to remove the flush water. The draining reduces the cesium concentration in the vessels by reducing the volume of cesium-containing material. The flushing, and subsequent draining, reduces the cesium in the vessels by diluting the cesium that remains in the film or drops on the vessel walls. MCU personnel requested that Savannah River National Laboratory (SRNL) researchers conduct a literature search to identify models to calculate the thickness of the liquid films remaining in the contactors, process tanks, and piping following draining of salt solution, solvent, and strip solution. The conclusions from this work are: (1) The predicted film thickness of the strip effluent is 0.010 mm on vertical walls, 0.57 mm on horizontal walls and 0.081 mm in horizontal pipes. (2) The predicted film thickness of the salt solution is 0.015 mm on vertical walls, 0.74 mm on horizontal walls, and 0.106 mm in horizontal pipes. (3) The predicted film thickness of the solvent is 0.022 mm on vertical walls, 0.91 mm on horizontal walls, and 0.13 mm in horizontal pipes. (4) The calculated film volume following draining is: (a) Salt solution receipt tank--1.6 gallons; (b) Salt solution feed tank--1.6 gallons; (c) Decontaminated salt solution hold tank--1.6 gallons; (d) Contactor drain tank--0.40 gallons; (e) Strip effluent hold tank--0.33 gallons; (f) Decontaminated salt solution decanter--0.37 gallons; (g) Strip effluent decanter--0.14 gallons; (h) Solvent hold tank--0.30 gallon; and (i) Corrugated piping between contactors--16-21 mL. (5) After the initial vessel draining, flushing the vessels with 100 gallons of water using a spray nozzle that produces complete vessel coverage and draining the flush water reduces the source term by the following amounts: (i) Salt solution receipt tank--63X; (ii) Salt solution feed tank--63X; (iii) Decontaminated salt solution hold tank--63X; (iv) Contactor drain tank--250X; (v) Strip effluent hold tank--300X; (vi) Decontaminated salt solution decanter--270X; (vii) Strip effluent decanter--710X; (viii) Solvent hold tank--330X. Understand that these estimates of film thickness are based on laboratory testing and fluid mechanics theory. The calculations assume drainage occurs by film flow. Much of the data used to develop the models came from tests with very ''clean'' fluids. Impurities in the fluids and contaminants on the vessels walls could increase liquid holdup. The application of film thickness models and source term reduction calculations should be considered along with operational conditions and H-Tank Farm/Liquid Waste operating experience. These calculations exclude the PVV/HVAC duct work and piping, as well as other areas that area outside the scope of this report.

Poirier, M; Fernando Fondeur, F; Samuel Fink, S

2006-06-06T23:59:59.000Z

303

Waste Characterization Data Manual for the inactive liquid low-level waste tank systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee. Environmental Restoration Program  

Science Conference Proceedings (OSTI)

This Waste Characterization Data Manual contains the results of an analysis of the contents of liquid low-level waste (LLLW) tanks that have been removed from service in accordance with the requirements of the Oak Ridge National Laboratory (ORNL) Federal Facility Agreement (FFA), Section IX.G.1. Section IX.G.1 of the FFA requires waste characterizations be conducted and provided to EPA and TDEC for all LLLW tanks that are removed from service. These waste characterizations shall include the results of sampling and analysis of the tank contents, including wastes, liquids, and sludges. This manual was first issued as ORNL/ER-80 in June 1992. The waste characterization data were extracted from ORNL reports that described tank sampling and analysis conducted in 1988 for 32 out-of-service tanks. This revision of the manual contains waste characterization data for 54 tanks, including the 32 tanks from the 1988 sampling campaign (Sects. 2.1 through 2.32) and the 22 additional tanks from a subsequent sampling campaign in 1992 and 1993 (Sects. 2.33 through 2.54). Data are presented from analyses of volatile organic compounds, semivolatile organic compounds, polychlorinated biphenyls (PCBs), pesticides, radiochemical compounds, and inorganic compounds. As additional data resulting from analyses of out-of-service tank samples become available, they will be added to this manual.

Not Available

1993-06-01T23:59:59.000Z

304

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

Science Conference Proceedings (OSTI)

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.

Howden, G.F.

1994-10-24T23:59:59.000Z

305

Off-gas characteristics of defense waste vitrification using liquid-fed Joule-heated ceramic melters  

DOE Green Energy (OSTI)

Off-gas and effluent characterization studies have been established as part of a PNL Liquid-Fed Ceramic Melter development program supporting the Savannah River Laboratory Defense Waste Processing Facility (SRL-DWPF). The objectives of these studies were to characterize the gaseous and airborne emission properties of liquid-fed joule-heated melters as a function of melter operational parameters and feed composition. All areas of off-gas interest and concern including effluent characterization, emission control, flow rate behavior and corrosion effects have been studied using alkaline and formic-acid based feed compositions. In addition, the behavioral patterns of gaseous emissions, the characteristics of melter-generated aerosols and the nature and magnitude of melter effluent losses have been established under a variety of feeding conditions with and without the use of auxiliary plenum heaters. The results of these studies have shown that particulate emissions are responsible for most radiologically important melter effluent losses. Melter-generated gases have been found to be potentially flammable as well as corrosive. Hydrogen and carbon monoxide present the greatest flammability hazard of the combustibles produced. Melter emissions of acidic volatile compounds of sulfur and the halogens have been responsible for extensive corrosion observed in melter plenums and in associated off-gas lines and processing equipment. The use of auxiliary plenum heating has had little effect upon melter off-gas characteristics other than reducing the concentrations of combustibles.

Goles, R.W.; Sevigny, G.J.

1983-09-01T23:59:59.000Z

306

Tank waste remediation system process engineering instruction manual  

SciTech Connect

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.

ADAMS, M.R.

1998-11-04T23:59:59.000Z

307

Modeling Coupled Processes in Clay Formations for Radioactive Waste Disposal  

Science Conference Proceedings (OSTI)

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.

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

2010-08-31T23:59:59.000Z

308

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

SciTech Connect

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.

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

2010-02-01T23:59:59.000Z

309

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

Science Conference Proceedings (OSTI)

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.

Not Available

1994-09-01T23:59:59.000Z

310

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

DOE Patents (OSTI)

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.

Silver, G.L.

1980-09-24T23:59:59.000Z

311

Compact Liquid Waste Evaporator for Cleanup on Hanfords Hot Cells [FULL PAPER  

SciTech Connect

Removal of radionuclide and hazardous contaminants from hot cells in Hanford's 324 Building will produce an aqueous waste stream requiring volume reduction and packaging. This paper describes a compact and remotely-operated evaporator system that was designed for use in the 324 Building's B-Cell (a shielded hot cell) to volume-reduce the waste waters that are generated from pressure washing of hot cell ceiling, wall, and floor surfaces. The evaporator incorporates an electric-heated reboiler to provide evaporation and drying to allow disposal of waste material. Design features of the evaporator system were strongly influenced by the need for remote handling and remote maintenance. Purified water vapor from the evaporation process will be released directly to the hot cell ventilation air.

HOBART, R.L.

2003-11-14T23:59:59.000Z

312

Characterization and monitoring of 300 Area Facility liquid waste streams: Status report  

SciTech Connect

This report summarizes the results of characterizing and monitoring the following sources during a portion of this year: liquid waste streams from Buildings 331, 320, and 3720; treated and untreated Columbia River water; and water at the confluence of the waste streams (that is, end-of-pipe). Characterization and monitoring data were evaluated for samples collected between March 22 and June 21, 1994, and subsequently analyzed for hazardous chemicals, radioactivity, and general parameters. Except for bis(2-ethylhexyl)phthalate, concentrations of chemicals detected and parameters measured at end-of-pipe were below the US Environmental Protection Agency existing and proposed drinking water standards. The source of the chemicals, except bis(2-ethylhexyl)phthalate, is not currently known. The bis(2-ethylhexyl)phthalate is probably an artifact of the plastic tubing used in the early stages of the sampling program. This practice was stopped. Concentrations and clearance times for contaminants at end-of-pipe depended strongly on source concentration at the facility release point, waste stream flow rates, dispersion, and the mechanical action of sumps. When present, the action of sumps had the greatest impact on contaminant clearance times. In the absence of sump activity, dispersion and flow rate were the controlling factors.

Manke, K.L. [ed.; Riley, R.G.; Ballinger, M.Y.; Damberg, E.G.; Evans, J.C.; Ikenberry, A.S.; Olsen, K.B.; Ozanich, R.M.; Thompson, C.J.

1994-09-01T23:59:59.000Z

313

Idaho Site Taps Old World Process to Treat Nuclear Waste | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Idaho Site Taps Old World Process to Treat Nuclear Waste Idaho Site Taps Old World Process to Treat Nuclear Waste Idaho Site Taps Old World Process to Treat Nuclear Waste September 9, 2013 - 12:00pm Addthis The Idaho site's sodium distillation system. The Idaho site's sodium distillation system. The top of a sodium distillation vessel, where waste enters the system. The top of a sodium distillation vessel, where waste enters the system. The Idaho site's sodium distillation system. The top of a sodium distillation vessel, where waste enters the system. IDAHO FALLS, Idaho - The EM program at the Idaho site is using an age-old process to treat transuranic (TRU) waste left over from nuclear reactor experiments. Developed in the first century and perfected by moonshiners in the 19th century, distillation will be used at the Idaho Nuclear Technology and

314

Conversion of MixAlco Process Sludge to Liquid Transportation Fuels  

E-Print Network (OSTI)

About 8 tons of dry undigested solid waste is generated by the MixAlco process for every 40 tons of food residue waste fed into the process. This MixAlco process produces liquid fuels and the sludge generated can be further converted into synthesis gas using the process of pyrolysis. The hydrogen component of the product synthesis gas may be separated by pressure swing adsorption and used in the hydrogenation of ketones into fuels and chemicals. The synthesis gas may also be catalytically converted into liquid fuels via the Fischer-Tropsch synthesis process. The auger-type pyrolyzer was operated at a temperature between 630-770 degrees C and at feed rates in the range of 280-374 g/minute. The response surface statistical method was used to obtain the highest syngas composition of 43.9 +/- 3.36 v % H2/33.3 +/- 3.29 v % CO at 740 degrees C. The CH4 concentration was 20.3 +/- 2.99 v %. For every ton of sludge pyrolyzed, 5,990 g H2 (719.3 MJ), 65,000 g CO (660 MJ) and 21,170 g CH4 (1055.4 MJ) were projected to be produced at optimum condition. At all temperatures, the sum of the energies of the products was greater than the electrical energy needed to sustain the process, making it energy neutral. To generate internal H2 for the MixAlco process, a method was developed to efficiently separate H2 using pressure swing adsorption (PSA) from the synthesis gas, with activated carbon and molecular sieve 5A as adsorbents. The H2 can be used to hydrogenate ketones generated from the MixAlco process to more liquid fuels. Breakthrough curves, cycle mass balances and cycle bed productivities (CBP) were used to determine the maximum hydrogen CBP using different adsorbent amounts at a synthesis gas feed rate of 10 standard lpm and pressure of 118 atm. A 99.9 % H2 purity was obtained. After a maximum CBP of 66 % was obtained further increases in % recovery led to a decrease in CBP. The synthesis gas can also be catalytically converted into liquid fuels by the Fischer-Tropsch synthesis (FTS) process. A Co-SiO2/Mo-Pd-Pt-ZSM-5 catalyst with a metal-metal-acid functionality was synthesized with the aim of increasing the selectivity of JP-8 (C10-C17) fuel range. The specific surface areas of the two catalysts were characterized using the BET technique. The electron probe microanalyzer (with WDS and EDS capabilities) was then used to confirm the presence of the applied metals Co, Mo, Pd and Pt on the respective supports. In addition to the gasoline (C4-C12) also produced, the synthesis gas H2:CO ratio was also adjusted to 1.90 for optimum cobalt performance in an enhanced FTS process. At 10 atm (150 psig) and 250 degrees C, the conventional FTS catalyst Co-SiO2 produced fuels rich in hydrocarbons within the gasoline carbon number range. At the same conditions the Co-SiO2-Mo-Pd-Pt/HZSM-5 catalyst increased the selectivity of JP-8. When Co-SiO2/Mo-Pd-Pt-HZSM-5 was used at 13.6 atm (200 psig) and 250 degrees C, a further increase in the selectivity of JP-8 and to some extent diesel was observed. The relative amounts of olefins and n-paraffins decreased with the products distribution shifting more towards the production of isomers.

Teiseh, Eliasu 1973-

2012-05-01T23:59:59.000Z

315

Using Ionic Liquids in Selective Hydrocarbon Conversion Processes  

DOE Green Energy (OSTI)

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.

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

316

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

SciTech Connect

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.

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

1981-10-01T23:59:59.000Z

317

Overview of Waste Processing & Recycling of Nonferrous Metals in ...  

Science Conference Proceedings (OSTI)

Selective Recovery of Gold from E-wastes by Using Cellulosic Wastes · Stabilization of Chromium-Based Slags with FeS2 and FeSO4 · Sulphide Precipitation ...

318

Recovery of Iron from Waste Slag of Pyrite Processing Using ...  

Science Conference Proceedings (OSTI)

Selective Recovery of Gold from E-wastes by Using Cellulosic Wastes · Stabilization of Chromium-Based Slags with FeS2 and FeSO4 · Sulphide Precipitation ...

319

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

SciTech Connect

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.

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

2008-06-25T23:59:59.000Z

320

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

SciTech Connect

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.

Fox, K.

2009-01-28T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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321

Unit Process Modeling [Nuclear Waste Management using Electrometallurg...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

322

Process to upgrade coal liquids by extraction prior to hydrodenitrogenation  

DOE Patents (OSTI)

Oxygen compounds are removed, e.g., by extraction, from a coal liquid prior to its hydrogenation. As a result, compared to hydrogenation of such a non-treated coal liquid, the rate of nitrogen removal is increased.

Schneider, Abraham (Overbrook Hills, PA); Hollstein, Elmer J. (Wilmington, DE); Janoski, Edward J. (Havertown, PA); Scheibel, Edward G. (Media, PA)

1982-01-01T23:59:59.000Z

323

Waste receiving and processing plant control system; system design description  

Science Conference Proceedings (OSTI)

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.

LANE, M.P.

1999-02-24T23:59:59.000Z

324

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

DOE Patents (OSTI)

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.

Kalb, P.D.; Colombo, P.

1997-07-15T23:59:59.000Z

325

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

DOE Patents (OSTI)

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.

Kalb, P.D.; Colombo, P.

1998-03-24T23:59:59.000Z

326

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

DOE Patents (OSTI)

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.

Kalb, Paul D. (Wading River, NY); Colombo, Peter (Patchogue, NY)

1999-07-20T23:59:59.000Z

327

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

DOE Patents (OSTI)

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.

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

328

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

DOE Patents (OSTI)

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.

Kalb, Paul D. (Wading River, NY); Colombo, Peter (Patchogue, NY)

1998-03-24T23:59:59.000Z

329

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

DOE Patents (OSTI)

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.

Kalb, P.D.; Colombo, P.

1999-07-20T23:59:59.000Z

330

Dynamic modeling and multivariable control of organic Rankine cycles in waste heat utilizing processes  

Science Conference Proceedings (OSTI)

In this paper, the dynamics of organic Rankine cycles (ORCs) in waste heat utilizing processes is investigated, and the physical model of a 100 kW waste heat utilizing process is established. In order to achieve both transient performance and steady-state ... Keywords: Linear quadratic regulator, Organic Rankine cycles, Process control

Jianhua Zhang; Wenfang Zhang; Guolian Hou; Fang Fang

2012-09-01T23:59:59.000Z

331

Microsoft PowerPoint - 1-07 Mason DOE EM Waste Processing Technology...  

NLE Websites -- All DOE Office Websites (Extended Search)

plants: Studsvik Processing Facility: Ion exchange resins (45" diameter FBSR) DOE Idaho Integrated Waste Treatment Unit: SBW treatment (48" diameter FBSR) DOE...

332

Nuclear criticality safety analysis summary report: The S-area defense waste processing facility  

SciTech Connect

The S-Area Defense Waste Processing Facility (DWPF) can process all of the high level radioactive wastes currently stored at the Savannah River Site with negligible risk of nuclear criticality. The characteristics which make the DWPF critically safe are: (1) abundance of neutron absorbers in the waste feeds; (2) and low concentration of fissionable material. This report documents the criticality safety arguments for the S-Area DWPF process as required by DOE orders to characterize and to justify the low potential for criticality. It documents that the nature of the waste feeds and the nature of the DWPF process chemistry preclude criticality.

Ha, B.C.

1994-10-21T23:59:59.000Z

333

GRR/Section 18 - Waste and Hazardous Material Assessment Process | Open  

Open Energy Info (EERE)

- Waste and Hazardous Material Assessment Process - Waste and Hazardous Material Assessment Process < GRR Jump to: navigation, search GRR-logo.png GEOTHERMAL REGULATORY ROADMAP Roadmap Home Roadmap Help List of Sections Section 18 - Waste and Hazardous Material Assessment Process 18 - WasteAndHazardousMaterialAssessmentProcess.pdf Click to View Fullscreen Contact Agencies Environmental Protection Agency Regulations & Policies RCRA CERCLA 40 CFR 261 Triggers None specified Click "Edit With Form" above to add content 18 - WasteAndHazardousMaterialAssessmentProcess.pdf Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Flowchart Narrative The use of underground and above ground storage tanks, discovery of waste

334

GRR/Section 18-AK-c - Waste Disposal Permit Process | Open Energy  

Open Energy Info (EERE)

AK-c - Waste Disposal Permit Process AK-c - Waste Disposal Permit Process < GRR Jump to: navigation, search GRR-logo.png GEOTHERMAL REGULATORY ROADMAP Roadmap Home Roadmap Help List of Sections Section 18-AK-c - Waste Disposal Permit Process 18AKC - WasteDisposalPermitProcess (1).pdf Click to View Fullscreen Contact Agencies Alaska Department of Environmental Conservation Regulations & Policies AS 46.03.110 Waste Disposal Permit Regulations 18 AAC 60.200 et seq Triggers None specified Click "Edit With Form" above to add content 18AKC - WasteDisposalPermitProcess (1).pdf Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Flowchart Narrative The Alaska Department of Environmental Conservation (DEC) is responsible

335

GRR/Section 18-OR-b - Hazardous Waste Permit Process | Open Energy  

Open Energy Info (EERE)

OR-b - Hazardous Waste Permit Process OR-b - Hazardous Waste Permit Process < GRR Jump to: navigation, search GRR-logo.png GEOTHERMAL REGULATORY ROADMAP Roadmap Home Roadmap Help List of Sections Section 18-OR-b - Hazardous Waste Permit Process 18ORBHazardousWastePermitProcess (1).pdf Click to View Fullscreen Contact Agencies United States Environmental Protection Agency Oregon Department of Environmental Quality Oregon Public Health Division Oregon Public Utility Commission Oregon Department of Fish and Wildlife Oregon Water Resources Department Regulations & Policies OAR 340-105: Management Facility Permits OAR 340-120: Hazardous Waste Management ORS 466: Storage, Treatment, and Disposal Triggers None specified Click "Edit With Form" above to add content 18ORBHazardousWastePermitProcess (1).pdf

336

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

E-Print Network (OSTI)

uniformity. Shredding of mixed waste to about 10 centimetersand untreated mixed municipal waste. GTZ/Holcim (2006) givesCEMBUREAU 2009). Mixed municipal waste must be pre-processed

Hasanbeigi, Ali

2013-01-01T23:59:59.000Z

337

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

E-Print Network (OSTI)

n.d. “Co-Processing of Waste and Energy Efficiency By CementAnnual North American Waste-to-Energy Conference NAWTEC17,2009. Stantec, 2011. Waste to Energy: a Technical Review of

Hasanbeigi, Ali

2013-01-01T23:59:59.000Z

338

Patterning Nanoparticle-Based Arrays through a Liquid Process  

Science Conference Proceedings (OSTI)

Author(s), Kathy Lu, Chase Hammond. On-Site Speaker (Planned), Kathy Lu ... Agricultural-Waste Biomass for Hydrogen Adsorption via Nano-Particle Synthesis

339

Processing of Discarded Liquid Crystal Display for Recovering Indium  

Science Conference Proceedings (OSTI)

Leaching Toxicity of Pb and Ba Containing in Cathode Ray Tube Glasses by SEP -TCLP · Mechanical Recycling of Electronic Wastes for Materials Recovery.

340

Waste Heat Recovery from Refrigeration in a Meat Processing Facility  

E-Print Network (OSTI)

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 recovery system consists of a shell and tube heat exchanger (16"? x 14'0") installed in the compressor hot gas discharge line. Water is recirculated from a 23,000-gallon tempered water storage tank to the heat exchanger by a circulating pump at the rate of 100 gallons per minute. All make-up water to the plant hot water system is supplied from this tempered water storage tank, which is maintained at a constant filled level. Tests to determine the actual rate of heat recovery were conducted from October 3, 1979 to October 12, 1979, disclosing an average usage of 147,000 gallons of hot water daily. These tests illustrated a varied heat recovery of from 0.5 to 1.0 million BTU per hour. The deviations were the result of both changing refrigeration demands and compressor operating modes. An average of 16 million BTU per day was realized, resulting in reduced boiler fuel costs of $30,000 annually, based on the present $.80 per gallon #2 fuel oil price. At the total installed cost of $79,000, including test instrumentation, the project was found to be economically viable. The study has demonstrated the technical and economic feasibility of refrigeration waste heat recovery as a positive energy conservation strategy which has broad applications in industry and commerce.

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

1980-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Preliminary analysis of the ORNL Liquid Low-Level Waste system  

Science Conference Proceedings (OSTI)

The objective of this report is to summarize the status of the Liquid Low-Level Waste (LLLW) Systems Analysis project. The focus of this project has been to collect and tabulate data concerning the LLLW system, analyze the current LLLW system operation, and develop the information necessary for the development of long-term treatment options for the LLLW generated at ORNL. The data used in this report were collected through a survey of Oak Ridge National Laboratory (ORNL) literature, various letter reports, and a survey of all current LLLW generators. These data are also being compiled in a user friendly database for ORNL-wide distribution. The database will allow the quick retrieval of all information collected on the ORNL LLLW system and will greatly benefit any LLLW analysis effort. This report summarizes the results for the analyses performed to date on the LLLW system.

Abraham, T.J.; DePaoli, S.M.; Robinson, S.M.; Walker, A.B.

1994-08-01T23:59:59.000Z

342

Applying the Systems Engineering Process for Establishing Requirements for the Safety and Health Monitoring System of the Waste Solidification Building at the Savannah River Site  

Science Conference Proceedings (OSTI)

The Safety and Health Monitoring (SHM) System technical basis document for the Waste Solidification Building (WSB) was developed by the Westinghouse Savannah River Company design team. The WSB is being designed and built to support the waste disposal needs of the Pit Disassembly and Conversion Facility (PDCF) and the Mixed Oxide Fuel Fabrication Facility (MFFF) at the Savannah River Site (SRS) in South Carolina. The main mission of the WSB is to process the radiological liquid waste streams from the PDCF and the MFFF into a solid waste form. The solid waste form, concrete encased waste, is acceptable for shipment and disposal as transuranic (TRU) waste at the Waste Isolation Pilot Plant (WIPP) and as Low Level Waste (LLW) at on-site disposal areas. The SHM System will also handle the job control waste from the PDCF, the MFFF, and the WSB. The SHM System will serve the WSB by monitoring personnel radiation exposure and environmental releases. The WSB design used HPT design support in determining the air monitoring equipment required for the WSB. The Systems Engineering (SE) process was applied to define the functions and requirements necessary to design and operate the SHM System. The SE process is a proven disciplined approach that supports management in clearly defining the mission or problem, managing system functions and requirements, identifying and managing risk, establishing bases for informed decision making, and verifying that products and services meet customer needs. This SE process applied to the SHM System was a major effort encompassing requirements analysis and interface control. Use of the SE process combined with HPT design input resulted in well-defined requirements to support the procurement of a safe-mission essential SHM System.

Simpkins, P.J.

2003-10-09T23:59:59.000Z

343

Followup of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process Systems Hazards Analysis Activity Review, March 2013  

NLE Websites -- All DOE Office Websites (Extended Search)

HSS Independent Activity Report - HSS Independent Activity Report - Rev. 0 Report Number: HIAR-WTP-2013-03-18 Site: Hanford Site Subject: Office of Enforcement and Oversight's Office of Safety and Emergency Management Evaluations Activity Report for Follow-up of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity Review Dates of Activity : 03/18/13 - 03/21/13 Report Preparer: James O. Low Activity Description/Purpose: The Office of Health, Safety and Security (HSS) staff observed a limited portion of the restart of the Hazard Analysis (HA) for the Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) Melter Process (LMP) System. The primary purpose of this HSS field activity, on March 18-21, 2013, was to observe and understand the revised approach

344

Followup of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process Systems Hazards Analysis Activity Review, March 2013  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

HSS Independent Activity Report - HSS Independent Activity Report - Rev. 0 Report Number: HIAR-WTP-2013-03-18 Site: Hanford Site Subject: Office of Enforcement and Oversight's Office of Safety and Emergency Management Evaluations Activity Report for Follow-up of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity Review Dates of Activity : 03/18/13 - 03/21/13 Report Preparer: James O. Low Activity Description/Purpose: The Office of Health, Safety and Security (HSS) staff observed a limited portion of the restart of the Hazard Analysis (HA) for the Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) Melter Process (LMP) System. The primary purpose of this HSS field activity, on March 18-21, 2013, was to observe and understand the revised approach

345

New desorption process treats refinery K and F wastes in demo trial  

SciTech Connect

A new desorption process for treating refinery wastes has been proven in pilot demonstrations at Amoco Oil Co.'s Texas City, Tex., refinery. The process -- Waste-Tech Services Inc.'s desorption and recovery unit (DRU) -- treats petroleum-contaminated refinery wastes and recovers oil and water suitable for recycling to the refinery. The DRU meets Resource Conservation and Recovery Act (RCRA) recycle exemptions and produces solids that satisfy US Environmental Protection Agency (EPA) land disposal restrictions (LDRs). This paper discusses RCRA wastes, the process, the demonstration unit, operating conditions, and analyses of semivolatiles, volatiles, leachable metals, and recovered oil and water.

Rasmussen, G.P. (Waste-Tech Services Inc., Golden, CO (United States))

1994-01-10T23:59:59.000Z

346

Electrochemical processing of nitrate waste solutions. Phase 2, Final report  

SciTech Connect

The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F{sup {minus}} ions from the synthetic mix migrating across the cation exchange membrane and forming HF in the acid anolyte. Other possibilities for anode materials were explored. A membrane separation process was investigated which employs an anion and cation exchange membrane to remove nitrite and nitrate, recovering caustic and nitric acid. Present research has shown poor current efficiencies for nitrite and nitrate transport across the anion exchange membrane due to co-migration of hydroxide anions. Precipitates form within the anion exchange membranes which would eventually result in the failure of the membranes. Electrochemical processing offers a highly promising and viable method for the treatment of nitrate waste solutions.

Genders, D.; Weinberg, N.; Hartsough, D. [Electrosynthesis Co., Inc., Cheektowaga, NY (US)

1992-10-07T23:59:59.000Z

347

Enhanced Chemical Cleaning: A New Process for Chemically Cleaning Savannah River Waste Tanks  

SciTech Connect

At the Savannah River Site (SRS) there are 49 High Level Waste (HLW) tanks that eventually must be emptied, cleaned, and closed. The current method of chemically cleaning SRS HLW tanks, commonly referred to as Bulk Oxalic Acid Cleaning (BOAC), requires about a half million liters (130,000 gallons) of 8 weight percent (wt%) oxalic acid to clean a single tank. During the cleaning, the oxalic acid acts as the solvent to digest sludge solids and insoluble salt solids, such that they can be suspended and pumped out of the tank. Because of the volume and concentration of acid used, a significant quantity of oxalate is added to the HLW process. This added oxalate significantly impacts downstream processing. In addition to the oxalate, the volume of liquid added competes for the limited available tank space. A search, therefore, was initiated for a new cleaning process. Using TRIZ (Teoriya Resheniya Izobretatelskikh Zadatch or roughly translated as the Theory of Inventive Problem Solving), Chemical Oxidation Reduction Decontamination with Ultraviolet Light (CORD-UV{reg_sign}), a mature technology used in the commercial nuclear power industry was identified as an alternate technology. Similar to BOAC, CORD-UV{reg_sign} also uses oxalic acid as the solvent to dissolve the metal (hydr)oxide solids. CORD-UV{reg_sign} is different, however, since it uses photo-oxidation (via peroxide/UV or ozone/UV to form hydroxyl radicals) to decompose the spent oxalate into carbon dioxide and water. Since the oxalate is decomposed and off-gassed, CORD-UV{reg_sign} would not have the negative downstream oxalate process impacts of BOAC. With the oxalate destruction occurring physically outside the HLW tank, re-precipitation and transfer of the solids, as well as regeneration of the cleaning solution can be performed without adding additional solids, or a significant volume of liquid to the process. With a draft of the pre-conceptual Enhanced Chemical Cleaning (ECC) flowsheet, taking full advantage of the many CORD-UV{reg_sign} benefits, performance demonstration testing was initiated using available SRS sludge simulant. The demonstration testing confirmed that ECC is a viable technology, as it can dissolve greater than 90% of the sludge simulant and destroy greater than 90% of the oxalates. Additional simulant and real waste testing are planned.

Ketusky, Edward; Spires, Renee; Davis, Neil

2009-02-11T23:59:59.000Z

348

Accident Fault Trees for Defense Waste Processing Facility  

Science Conference Proceedings (OSTI)

The purpose of this report is to document fault tree analyses which have been completed for the Defense Waste Processing Facility (DWPF) safety analysis. Logic models for equipment failures and human error combinations that could lead to flammable gas explosions in various process tanks, or failure of critical support systems were developed for internal initiating events and for earthquakes. These fault trees provide frequency estimates for support systems failures and accidents that could lead to radioactive and hazardous chemical releases both on-site and off-site. Top event frequency results from these fault trees will be used in further APET analyses to calculate accident risk associated with DWPF facility operations. This report lists and explains important underlying assumptions, provides references for failure data sources, and briefly describes the fault tree method used. Specific commitments from DWPF to provide new procedural/administrative controls or system design changes are listed in the ''Facility Commitments'' section. The purpose of the ''Assumptions'' section is to clarify the basis for fault tree modeling, and is not necessarily a list of items required to be protected by Technical Safety Requirements (TSRs).

Sarrack, A.G.

1999-06-22T23:59:59.000Z

349

Understanding Cement Waste Forms  

Science Conference Proceedings (OSTI)

Oct 29, 2009 ... Ongoing nuclear operations, decontamination and decommissioning, salt waste disposal, and closure of liquid waste tanks result in ...

350

Waste Receiving and Processing Facility Module 1 Data Management System Software Requirements Specification  

Science Conference Proceedings (OSTI)

This document provides the software requirements for Waste Receiving and Processing (WRAP) Module 1 Data Management System (DMS). The DMS is one of the plant computer systems for the new WRAP 1 facility (Project W-026). The DMS will collect, store and report data required to certify the low level waste (LLW) and transuranic (TRU) waste items processed at WRAP 1 as acceptable for shipment, storage, or disposal.

Brann, E.C. II

1994-09-09T23:59:59.000Z

351

CERTIFICATION DOCKET FOR THE F0RhqE.R SITE OF THE RADIOACTIVE LIQUID WASTE TREATMENT PLANT (TA-45)  

Office of Legacy Management (LM)

CERTIFICATION DOCKET CERTIFICATION DOCKET FOR THE F0RhqE.R SITE OF THE RADIOACTIVE LIQUID WASTE TREATMENT PLANT (TA-45) AND THE EFFLUENT RECEIVING AREAS OF ACID, PUEBLO, AND LOS ALAMOS CANYOM, LOS ALAMOS, NEW MEXICO DEPARTMENT OF ENERGY Office of Nuclear Energy Office of Terminal Waste Disposal and Remedial Action Division of Remedial Action Projects -. CONTENTS A Page - Introduction to the Certification Docket for the Former Site of the Radioactive Liquid Waste Treatment Plant (TA-45) and the Effluent Receiving Areas of Acid, Pueblo, and Los Alamos Canyons, Los Alamos, New Mexico Description of the Formeriy Utilized Sites Program at the Former Site of the T.4-45 Treatment Plant and Acid, Pueblo, and Los Alamos Canyons Purpose Property Identification Docket Contents

352

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

Science Conference Proceedings (OSTI)

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.

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

2012-01-09T23:59:59.000Z

353

Characterization and monitoring of 300 Area facility liquid waste streams: 1994 Annual report  

Science Conference Proceedings (OSTI)

This report summarizes the results of characterizing and monitoring the following sources during calendar year 1994: liquid waste streams from Buildings 306, 320, 324, 326, 331, and 3720 in the 300 Area of Hanford Site and managed by the Pacific Northwest Laboratory; treated and untreated Columbia River water (influent); and water at the confluence of the waste streams (that is, end-of-pipe). Data were collected from March to December before the sampling system installation was completed. Data from this initial part of the program are considered tentative. Samples collected were analyzed for chemicals, radioactivity, and general parameters. In general, the concentrations of chemical and radiological constituents and parameters in building wastewaters which were sampled and analyzed during CY 1994 were similar to historical data. Exceptions were the occasional observances of high concentrations of chloride, nitrate, and sodium that are believed to be associated with excursions that were occurring when the samples were collected. Occasional observances of high concentrations of a few solvents also appeared to be associated with infrequent building r eases. During calendar year 1994, nitrate, aluminum, copper, lead, zinc, bis(2-ethylhexyl) phthalate, and gross beta exceeded US Environmental Protection Agency maximum contaminant levels.

Riley, R.G.; Ballinger, M.Y.; Damberg, E.G.; Evans, J.C.; Julya, J.L.; Olsen, K.B.; Ozanich, R.M.; Thompson, C.J.; Vogel, H.R.

1995-04-01T23:59:59.000Z

354

Chapter 38 Hazardous Waste Permitting Process (Kentucky) | Open...  

Open Energy Info (EERE)

Policy Contact Contact Name Anthony Hatton (director) Department Department for Environmental Protection Division Division of Waste Management Address 200 Fair Oaks L.,...

355

Metals Mobilization During E-Waste Bioleaching Process: Effect of ...  

Science Conference Proceedings (OSTI)

A Study on Waste Packaging Containers Generated by Household in Taiwan ... Mullites Bodies Produced From the Kaolin Residue Using Microwave Energy.

356

Innovations through Ceramic Processing by Tailoring Solid-Liquid ...  

Science Conference Proceedings (OSTI)

Abstract Scope, Tailoring the solid-gas and solid-liquid interfaces of particles by ... By using principles found in natural composites, layered polymer/ceramic ...

357

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

Science Conference Proceedings (OSTI)

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.

Not Available

1994-11-01T23:59:59.000Z

358

Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Oversight Assessment of Oversight Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project May 2011 January 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Independent Oversight Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project

359

Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project  

NLE Websites -- All DOE Office Websites (Extended Search)

Oversight Assessment of Oversight Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project May 2011 January 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Independent Oversight Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project

360

Temperature programmed combustion studies of the co-processing of coal and waste materials  

E-Print Network (OSTI)

Temperature programmed combustion studies of the co-processing of coal and waste materials F) to study the interaction between coal, polyethylene, and dried sewage sludge which are possible components in coal/ waste materials co-processing combustion systems. The TPC studies were carried out on the raw

Thomas, Mark

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

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

Science Conference Proceedings (OSTI)

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.

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

362

SRS - Programs - Waste Solidification  

NLE Websites -- All DOE Office Websites (Extended Search)

Waste Solidification Waste Solidification The two primary facilities operated within the Waste Solidification program are Saltstone and the Defense Waste Processing Facility (DWPF). Each DWPF canister is 10 feet tall and 2 feet in diameter, and typically takes a little over a day to fill. Each DWPF canister is 10 feet tall and 2 feet in diameter, and typically takes a little over a day to fill. The largest radioactive waste glassification plant in the world, DWPF converts the high-level liquid nuclear waste currently stored at the Savannah River Site (SRS) into a solid glass form suitable for long-term storage and disposal. Scientists have long considered this glassification process, called "vitrification," as the preferred option for immobilizing high-level radioactive liquids into a more stable, manageable form until a federal

363

Online elemental analysis of process gases with ICP-OES: A case study on waste wood combustion  

SciTech Connect

Highlights: Black-Right-Pointing-Pointer Simultaneous measurements of 23 elements in process gases of a waste wood combustor. Black-Right-Pointing-Pointer Mobile ICP spectrometer allows measurements of high quality at industrial plants. Black-Right-Pointing-Pointer Continuous online measurements with high temporal resolution. Black-Right-Pointing-Pointer Linear correlations among element concentrations in the raw flue gas were detected. Black-Right-Pointing-Pointer Novel sampling and calibration methods for ICP-OES analysis of process gases. - Abstract: A mobile sampling and measurement system for the analysis of gaseous and liquid samples in the field was developed. An inductively coupled plasma optical emission spectrometer (ICP-OES), which is built into a van, was used as detector. The analytical system was calibrated with liquid and/or gaseous standards. It was shown that identical mass flows of either gaseous or liquid standards resulted in identical ICP-OES signal intensities. In a field measurement campaign trace and minor elements in the raw flue gas of a waste wood combustor were monitored. Sampling was performed with a highly transport efficient liquid quench system, which allowed to observe temporal variations in the elemental process gas composition. After a change in feedstock an immediate change of the element concentrations in the flue gas was detected. A comparison of the average element concentrations during the combustion of the two feedstocks showed a high reproducibility for matrix elements that are expected to be present in similar concentrations. On the other hand elements that showed strong differences in their concentration in the feedstock were also represented by a higher concentration in the flue gas. Following the temporal variations of different elements revealed strong correlations between a number of elements, such as chlorine with sodium, potassium and zinc, as well as arsenic with lead, and calcium with strontium.

Wellinger, Marco, E-mail: marco.wellinger@gmail.com [General Energy Research Department, Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), School of Architecture, Civil and Environmental Engineering (ENAC-IIE), CH-1015 Lausanne (Switzerland); Wochele, Joerg; Biollaz, Serge M.A. [General Energy Research Department, Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ludwig, Christian, E-mail: christian.ludwig@psi.ch [General Energy Research Department, Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), School of Architecture, Civil and Environmental Engineering (ENAC-IIE), CH-1015 Lausanne (Switzerland)

2012-10-15T23:59:59.000Z

364

GRR/Section 18-AK-b - Hazardous Waste Permit Process | Open Energy  

Open Energy Info (EERE)

8-AK-b - Hazardous Waste Permit Process 8-AK-b - Hazardous Waste Permit Process < GRR Jump to: navigation, search GRR-logo.png GEOTHERMAL REGULATORY ROADMAP Roadmap Home Roadmap Help List of Sections Section 18-AK-b - Hazardous Waste Permit Process 18AKB - HazardousWastePermitProcess (1).pdf Click to View Fullscreen Contact Agencies Alaska Department of Environmental Conservation United States Environmental Protection Agency Regulations & Policies AS 46.03.302 18 AAC 60.020 Triggers None specified Click "Edit With Form" above to add content 18AKB - HazardousWastePermitProcess (1).pdf Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Flowchart Narrative The Alaska Department of Environmental Conservation defers to the federal

365

Process and material that encapsulates solid hazardous waste  

DOE Patents (OSTI)

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.

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

1999-01-01T23:59:59.000Z

366

Recycling of Wastes Generated during the Steelmaking Process  

Science Conference Proceedings (OSTI)

These are wastes with considerable production and limited applications, therefore this work studied the recovery of these wastes into ... Clayey Ceramic Incorporated with Powder from the Sintering Plant of a Steel-Making Industry ... Influence of Fly Ash and Fluorgypsum on Hydration Heat and Mortar Strength of Cement.

367

Process and material that encapsulates solid hazardous waste  

DOE Patents (OSTI)

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

O' Brien, Michael H.; Erickson, Arnold W.

1997-12-01T23:59:59.000Z

368

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)

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

Demirel, Melik C.

369

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

SciTech Connect

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.

Not Available

2011-10-01T23:59:59.000Z

370

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

SciTech Connect

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.

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

2002-07-18T23:59:59.000Z

371

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

SciTech Connect

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.

Nichols, Todd Travis; Taylor, Dean Dalton

2002-07-01T23:59:59.000Z

372

Thermal Processing of Wastes - III (Eaf and Iron & Steel Plant Wastes)  

Science Conference Proceedings (OSTI)

... OF WASTE MATERIALS IN A FERROCHROME INDUSTRY: S. A. Platias, MIRTEC S. A., Research Division, Á Industrial Area, 38500 Volos, Greece. 2:50 pm.

373

Glovebox design requirements for molten salt oxidation processing of transuranic waste  

SciTech Connect

This paper presents an overview of potential technologies for stabilization of {sup 238}Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from {sup 238}Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented.

Ramsey, K.B.; Acosta, S.V. [Los Alamos National Lab., NM (United States); Wernly, K.D. [Molten Salt Oxidation Corp., Bensalem, PA (United States)

1998-12-31T23:59:59.000Z

374

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

Science Conference Proceedings (OSTI)

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.

S. K. Evans

2006-08-15T23:59:59.000Z

375

Nuclear Solid Waste Processing Design at the Idaho Spent Fuels Facility  

Science Conference Proceedings (OSTI)

A spent nuclear fuels (SNF) repackaging and storage facility was designed for the Idaho National Engineering and Environmental Laboratory (INEEL), with nuclear solid waste processing capability. Nuclear solid waste included contaminated or potentially contaminated spent fuel containers, associated hardware, machinery parts, light bulbs, tools, PPE, rags, swabs, tarps, weld rod, and HEPA filters. Design of the nuclear solid waste processing facilities included consideration of contractual, regulatory, ALARA (as low as reasonably achievable) exposure, economic, logistical, and space availability requirements. The design also included non-attended transfer methods between the fuel packaging area (FPA) (hot cell) and the waste processing area. A monitoring system was designed for use within the FPA of the facility, to pre-screen the most potentially contaminated fuel canister waste materials, according to contact- or non-contact-handled capability. Fuel canister waste materials which are not able to be contact-handled after attempted decontamination will be processed remotely and packaged within the FPA. Noncontact- handled materials processing includes size-reduction, as required to fit into INEEL permitted containers which will provide sufficient additional shielding to allow contact handling within the waste areas of the facility. The current design, which satisfied all of the requirements, employs mostly simple equipment and requires minimal use of customized components. The waste processing operation also minimizes operator exposure and operator attendance for equipment maintenance. Recently, discussions with the INEEL indicate that large canister waste materials can possibly be shipped to the burial facility without size-reduction. New waste containers would have to be designed to meet the drop tests required for transportation packages. The SNF waste processing facilities could then be highly simplified, resulting in capital equipment cost savings, operational time savings, and significantly improved ALARA exposure.

Dippre, M. A.

2003-02-25T23:59:59.000Z

376

Corrosion and failure processes in high-level waste tanks  

Science Conference Proceedings (OSTI)

A large amount of radioactive waste has been stored safely at the Savannah River and Hanford sites over the past 46 years. The aim of this report is to review the experimental corrosion studies at Savannah River and Hanford with the intention of identifying the types and rates of corrosion encountered and indicate how these data contribute to tank failure predictions. The compositions of the High-Level Wastes, mild steels used in the construction of the waste tanks and degradation-modes particularly stress corrosion cracking and pitting are discussed. Current concerns at the Hanford Site are highlighted.

Mahidhara, R.K.; Elleman, T.S.; Murty, K.L. [North Carolina State Univ., Raleigh, NC (United States)

1992-11-01T23:59:59.000Z

377

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

E-Print Network (OSTI)

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

Skogestad, Sigurd

378

Caustic Recycle from Hanford Tank Waste Using NaSICON Ceramic Membrane Salt Splitting Process  

Science Conference Proceedings (OSTI)

A family of inorganic ceramic materials, called sodium (Na) Super Ion Conductors (NaSICON), has been studied at Pacific Northwest National Laboratory (PNNL) to investigate their ability to separate sodium from radioactively contaminated sodium salt solutions for treating U.S. Department of Energy (DOE) tank wastes. Ceramatec Inc. developed and fabricated a membrane containing a proprietary NAS-GY material formulation that was electrochemically tested in a bench-scale apparatus with both a simulant and a radioactive tank-waste solution to determine the membrane performance when removing sodium from DOE tank wastes. Implementing this sodium separation process can result in significant cost savings by reducing the disposal volume of low-activity wastes and by producing a NaOH feedstock product for recycle into waste treatment processes such as sludge leaching, regenerating ion exchange resins, inhibiting corrosion in carbon-steel tanks, or retrieving tank wastes.

Fountain, Matthew S.; Kurath, Dean E.; Sevigny, Gary J.; Poloski, Adam P.; Pendleton, J.; Balagopal, S.; Quist, M.; Clay, D.

2009-02-20T23:59:59.000Z

379

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

Science Conference Proceedings (OSTI)

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.

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

1994-03-01T23:59:59.000Z

380

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

SciTech Connect

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.

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

2002-02-26T23:59:59.000Z

Note: This page contains sample records for the topic "liquid waste process" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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381

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

DOE Green Energy (OSTI)

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

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

1998-06-04T23:59:59.000Z

382

Proceedings: Vitrification of Low-Level Waste--the Process and Potential  

Science Conference Proceedings (OSTI)

Vitrification technology, or the consolidation of waste in a glass matrix, represents a proven method for achieving volume reduction for high-level industrial waste. Application of this technology is emerging as a viable treatment of low-level waste. This workshop focused on the range of vitrification technologies now available and highlighted issues associated with application of the vitrification process in the nuclear power industry.

1996-05-21T23:59:59.000Z

383

DOE Office of Waste Processing Technical Exchange - Agenda  

Discussion of Future R&D Needs: All-3:45. ID Calcine Treatment Options Study: Hagers: DOE-ID: 4:15: Waste Immobilization Community of Practice : Peeler: WSRC: RETURN ...

384

Process for immobilizing plutonium into vitreous ceramic waste forms  

DOE Patents (OSTI)

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.

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

1997-01-01T23:59:59.000Z

385

Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel  

SciTech Connect

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.

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

386

TRUEX process - a process for the extraction of the transuranic elements from nitric acid wastes utilizing modified PUREX solvent  

SciTech Connect

A generic transuranic (TRU) element extraction/recovery process was developed based on the use of octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide, O phi D(iB)CMPO, dissolved in PUREX process solvent (tributyl phosphate, TBP, in normal paraffinic hydrocarbon, NPH). The process (called TRUEX) is capable of reducing the TRU concentration by many orders of magnitude in waste solutions containing a wide range of nitric acid, salt, and fission product concentrations. A major feature of the process is that it is readily adaptable for waste processing in existing fuel reprocessing facilities.

Horwitz, E.P.; Kalina, D.G.; Diamond, H.; Vandegrift, G.F.; Schulz, W.W.

1985-01-01T23:59:59.000Z

387

Low temperature hydrothermal processing of organic contaminants in Hanford tank waste  

DOE Green Energy (OSTI)

Batch and continuous flow reactor tests at Pacific Northwest Laboratory (PNL) have shown that organics similar to those present in the single-shell and double-shell underground storage tanks at Hanford can be decomposed in the liquid phase at relatively mild temperatures of 150[degree]C to 350[degree]C in an aqueous process known as hydrothermal processing (HTP). The organics will react with the abundant oxidants such s nitrite already present in the Hanford tank waste to form hydrogen, carbon dioxide, methane, and ammonia. No air or oxygen needs to be added to the system. Ferrocyanides and free cyanide will hydrolyze at similar temperatures to produce formate and ammonia and may also react with nitrates or other oxides. During testing, the organic carbon was transformed first to oxalate at[approximately]310[degree]C and completely oxidized to carbonate at [approximately]350[degree]C accompanied by hydroxide consumption. Solids were formed at higher temperatures, causing a small-diameter outlet tube to plug. The propensity for plugging was reduced by diluting the feed with concentrated hydroxide.

Jones, E.O.; Pederson, L.R.; Freeman, H.D.; Schmidt, A.J. (Pacific Northwest Lab., Richland, WA (United States)); Babad, H. (Westinghouse Hanford Co., Richland, WA (United States))

1993-02-01T23:59:59.000Z

388

Low temperature hydrothermal processing of organic contaminants in Hanford tank waste  

DOE Green Energy (OSTI)

Batch and continuous flow reactor tests at Pacific Northwest Laboratory (PNL) have shown that organics similar to those present in the single-shell and double-shell underground storage tanks at Hanford can be decomposed in the liquid phase at relatively mild temperatures of 150{degree}C to 350{degree}C in an aqueous process known as hydrothermal processing (HTP). The organics will react with the abundant oxidants such s nitrite already present in the Hanford tank waste to form hydrogen, carbon dioxide, methane, and ammonia. No air or oxygen needs to be added to the system. Ferrocyanides and free cyanide will hydrolyze at similar temperatures to produce formate and ammonia and may also react with nitrates or other oxides. During testing, the organic carbon was transformed first to oxalate at{approximately}310{degree}C and completely oxidized to carbonate at {approximately}350{degree}C accompanied by hydroxide consumption. Solids were formed at higher temperatures, causing a small-diameter outlet tube to plug. The propensity for plugging was reduced by diluting the feed with concentrated hydroxide.

Jones, E.O.; Pederson, L.R.; Freeman, H.D.; Schmidt, A.J. [Pacific Northwest Lab., Richland, WA (United States); Babad, H. [Westinghouse Hanford Co., Richland, WA (United States)

1993-02-01T23:59:59.000Z

389

Facility design philosophy: Tank Waste Remediation System Process support and infrastructure definition  

Science Conference Proceedings (OSTI)

This report documents the current facility design philosophy for the Tank Waste Remediation System (TWRS) process support and infrastructure definition. The Tank Waste Remediation System Facility Configuration Study (FCS) initially documented the identification and definition of support functions and infrastructure essential to the TWRS processing mission. Since the issuance of the FCS, the Westinghouse Hanford Company (WHC) has proceeded to develop information and requirements essential for the technical definition of the TWRS treatment processing programs.

Leach, C.E.; Galbraith, J.D. [Westinghouse Hanford Co., Richland, WA (United States); Grant, P.R.; Francuz, D.J.; Schroeder, P.J. [Fluor Daniel, Inc., Richland, WA (United States)

1995-11-01T23:59:59.000Z

390

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

Science Conference Proceedings (OSTI)

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.

William Linak

2004-12-16T23:59:59.000Z

391

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

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.

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

392

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

SciTech Connect

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.

Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.

2011-09-12T23:59:59.000Z

393

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

SciTech Connect

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.

Bush, S.

2009-11-05T23:59:59.000Z

394

Demonstrating Reliable High Level Waste Slurry Sampling Techniques to Support Hanford Waste Processing - 14194  

SciTech Connect

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.

Kelly, Steven E.

2013-11-11T23:59:59.000Z

395

Information; and Other Matters- Amount of Uranium in Liquid Waste Effluents, Treated Domestic Sanitary Wastewater Sampling, and Liquid Effluent Collection and  

E-Print Network (OSTI)

for the AES exemption request related to commencement of construction (Ref. 2). On October 15, 2009, AES submitted the response to the NRC RAIs related to commencement of construction (Ref. 3). Subsequently, the NRC requested additional information regarding the AES response. Enclosure 1.1 provides the AES response to the additional information regarding preconstrucion activities requested by the NRC. Enclosure 2.1 provides the markup pages of the EREF ER. On August 10, 2009, the NRC transmitted to AES RAIs regarding the EREF Environmental Report (ER) (Ref. 4). On September 9, 2009, AES submitted the response to the NRC ER RAIs (Ref. 5). Subsequently, the NRC requested additional information regarding other matters including the amount of uranium in liquid waste effluents, treated domestic sanitary wastewater sampling, and Liquid Effluent Collection and Treatment System evaporator sediment sampling. Enclosure 1.2 provides the AES response regarding the amount of uranium in liquid waste effluent. There are no markup pages to the EREF ER for this response. Enclosure 1.3 provides the AES response regarding treated domestic sanitary wastewater sampling. Enclosure 2.2 provides the markup pages of the EREF ER. Enclosure 1.4 provides the AES

Eagle Rock; Enrichment Facility

2009-01-01T23:59:59.000Z

396

Optimal evaluation of infectious medical waste disposal companies using the fuzzy analytic hierarchy process  

SciTech Connect

Ever since Taiwan's National Health Insurance implemented the diagnosis-related groups payment system in January 2010, hospital income has declined. Therefore, to meet their medical waste disposal needs, hospitals seek suppliers that provide high-quality services at a low cost. The enactment of the Waste Disposal Act in 1974 had facilitated some improvement in the management of waste disposal. However, since the implementation of the National Health Insurance program, the amount of medical waste from disposable medical products has been increasing. Further, of all the hazardous waste types, the amount of infectious medical waste has increased at the fastest rate. This is because of the increase in the number of items considered as infectious waste by the Environmental Protection Administration. The present study used two important findings from previous studies to determine the critical evaluation criteria for selecting infectious medical waste disposal firms. It employed the fuzzy analytic hierarchy process to set the objective weights of the evaluation criteria and select the optimal infectious medical waste disposal firm through calculation and sorting. The aim was to propose a method of evaluation with which medical and health care institutions could objectively and systematically choose appropriate infectious medical waste disposal firms.

Ho, Chao Chung, E-mail: ho919@pchome.com.tw [Department of Industrial Management, National Taiwan University of Science and Technology, Taipei, Taiwan (China)

2011-07-15T23:59:59.000Z

397

Formulation and Characterization of Waste Glasses with Varying Processing Temperature  

Science Conference Proceedings (OSTI)

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.

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

398

Process for converting coal into liquid fuel and metallurgical coke  

DOE Patents (OSTI)

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.

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

1994-01-01T23:59:59.000Z

399

Nuclear Safety R&D in the Waste Processing Technology Development & Deployment Program  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

R&D in the Waste Processing R&D in the Waste Processing Technology Development & Deployment Program Presentation to the DOE High Level Waste Corporate Board July 29, 2009 Al Baione Office of