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1

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

Broader source: Energy.gov [DOE]

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

2

Materials evaluation programs at the Defense Waste Processing Facility  

SciTech Connect (OSTI)

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

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

1992-01-01T23:59:59.000Z

3

Materials evaluation programs at the Defense Waste Processing Facility  

SciTech Connect (OSTI)

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

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

1992-12-31T23:59:59.000Z

4

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

5

Defense Waste Processing Facility wasteform and canister description: Revision 2  

SciTech Connect (OSTI)

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

Baxter, R.G.

1988-12-01T23:59:59.000Z

6

Accident Fault Trees for Defense Waste Processing Facility  

SciTech Connect (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

7

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

SciTech Connect (OSTI)

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

8

Retrieval process development and enhancements waste simulant compositions and defensibility  

SciTech Connect (OSTI)

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

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

1997-09-01T23:59:59.000Z

9

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

10

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

SciTech Connect (OSTI)

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

Baxter, R G

1986-01-01T23:59:59.000Z

11

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

SciTech Connect (OSTI)

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

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

2008-01-15T23:59:59.000Z

12

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

SciTech Connect (OSTI)

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

Landon, L.F. (comp.)

1980-05-01T23:59:59.000Z

13

Checkout and start-up of the integrated DWPF (Defense Waste Processing Facility) melter system  

SciTech Connect (OSTI)

The Integrated DWPF Melter System (IDMS) is a one-ninth-scale demonstration of the Defense Waste Processing Facility (DWPF) feed preparation, melter, and off-gas systems. The IDMS will be the first engineering-scale melter system at SRL to process mercury and flowsheet levels of halides and sulfates. This report includes a summary of the IDMS program objectives, system and equipment descriptions, and detailed discussions of the system checkout and start-up. 10 refs., 44 figs., 20 tabs.

Smith, M.E.; Hutson, N.D.; Miller, D.H.; Morrison, J.; Shah, H.; Shuford, J.A.; Glascock, J.; Wurzinger, F.H.; Zamecnik, J.R.

1989-11-11T23:59:59.000Z

14

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

SciTech Connect (OSTI)

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

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

2013-01-09T23:59:59.000Z

15

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

16

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

SciTech Connect (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

17

Discrete event simulation of the Defense Waste Processing Facility (DWPF) analytical laboratory  

SciTech Connect (OSTI)

A discrete event simulation of the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) analytical laboratory has been constructed in the GPSS language. It was used to estimate laboratory analysis times at process analytical hold points and to study the effect of sample number on those times. Typical results are presented for three different simultaneous representing increasing levels of complexity, and for different sampling schemes. Example equipment utilization time plots are also included. SRS DWPF laboratory management and chemists found the simulations very useful for resource and schedule planning.

Shanahan, K.L.

1992-02-01T23:59:59.000Z

18

Erosion/corrosion concerns in feed preparation systems at the Defense Waste Processing Facility  

SciTech Connect (OSTI)

The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950`s to produce nuclear materials in support of the national defense effort. The Department of Energy authorized the construction of the Defense Waste Processing Facility (DWPF) to immobilize the high level radioactive waste resulting from these processes as a durable borosilicate glass. The DWPF, after having undergone extensive testing, has been approved for operations and is currently immobilizing radioactive waste. To ensure reliability of the DWPF remote canyon processing equipment, a materials evaluation program was performed prior to radioactive operations to determine to what extent erosion/corrosion would impact design life of equipment. The program consisted of performing pre-service baseline inspections on critical equipment and follow-up inspections after completion of DWPF cold chemical demonstration runs. Non-destructive examination (NDE) techniques were used to assess erosion/corrosion as well as evaluation of corrosion coupon racks. These results were used to arrive at predicted equipment life for selected feed preparation equipment. It was concluded with the exception of the coil and agitator for the slurry mix evaporator (SME), which are exposed to erosive glass frit particles, all of the equipment should meet its design life.

Gee, J.T.; Chandler, C.T.; Daugherty, W.L.; Imrich, K.J.; Jenkins, C.F.

1997-12-31T23:59:59.000Z

19

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

SciTech Connect (OSTI)

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

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

2012-09-14T23:59:59.000Z

20

DWPF (Defense Waste Processing Facility) canister impact testing and analyses for the Transportation Technology Center  

SciTech Connect (OSTI)

A legal weight truck cask design has been developed for the US Department of Energy by GA Technologies, Inc. The cask will be used to transport defense high-level waste canisters produced by the Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The development of the cask required the collection of impact data for the DWPF canisters. The Materials Characterization Center (MCC) performed this work under the guidance of the Transportation Technology Center (TTC) at Sandia National Laboratories. Two full-scale DWPF canisters filled with nonradioactive borosilicate glass were impacted under ''normal'' and ''hypothetical'' accident conditions. Two canisters, supplied by the DWPF, were tested. Each canister was vertically dropped on the bottom end from a height of either 0.3 m or 9.1 m (for normal or hypothetical accident conditions, respectively). The structural integrity of each canister was then examined using helium leak and dye penetrant testing. The canisters' diameters and heights, which had been previously measured, were then remeasured to determine how the canister dimensions had changed. Following structural integrity testing, the canisters were flaw leak tested. For transportation flaw leak testing, four holes were fabricated into the shell of canister A-27 (0.3 m drop height). The canister was then transported a total distance of 2069 miles. During transport, the waste form material that fell from each flaw was collected to determine the amount of size distribution of each flaw release. 2 refs., 8 figs., 12 tabs.

Farnsworth, R.K.; Mishima, J.

1988-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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

IMPACTS OF ANTIFOAM ADDITIONS AND ARGON BUBBLING ON DEFENSE WASTE PROCESSING FACILITY REDUCTION/OXIDATION  

SciTech Connect (OSTI)

During melting of HLW glass, the REDOX of the melt pool cannot be measured. Therefore, the Fe{sup +2}/{Sigma}Fe ratio in the glass poured from the melter must be related to melter feed organic and oxidant concentrations to ensure production of a high quality glass without impacting production rate (e.g., foaming) or melter life (e.g., metal formation and accumulation). A production facility such as the Defense Waste Processing Facility (DWPF) cannot wait until the melt or waste glass has been made to assess its acceptability, since by then no further changes to the glass composition and acceptability are possible. therefore, the acceptability decision is made on the upstream process, rather than on the downstream melt or glass product. That is, it is based on 'feed foward' statistical process control (SPC) rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. Use of the DWPF REDOX model has controlled the balanjce of feed reductants and oxidants in the Sludge Receipt and Adjustment Tank (SRAT). Once the alkali/alkaline earth salts (both reduced and oxidized) are formed during reflux in the SRAT, the REDOX can only change if (1) additional reductants or oxidants are added to the SRAT, the Slurry Mix Evaporator (SME), or the Melter Feed Tank (MFT) or (2) if the melt pool is bubble dwith an oxidizing gas or sparging gas that imposes a different REDOX target than the chemical balance set during reflux in the SRAT.

Jantzen, C.; Johnson, F.

2012-06-05T23:59:59.000Z

22

ANION ANALYSES BY ION CHROMATOGRAPHY FOR THE ALTERNATE REDUCTANT DEMONSTRATION FOR THE DEFENSE WASTE PROCESSING FACILITY  

SciTech Connect (OSTI)

The Process Science Analytical Laboratory (PSAL) at the Savannah River National Laboratory was requested by the Defense Waste Processing Facility (DWPF) to develop and demonstrate an Ion Chromatography (IC) method for the analysis of glycolate, in addition to eight other anions (fluoride, formate, chloride, nitrite, nitrate, sulfate, oxalate and phosphate) in Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) samples. The method will be used to analyze anions for samples generated from the Alternate Reductant Demonstrations to be performed for the DWPF at the Aiken County Technology Laboratory (ACTL). The method is specific to the characterization of anions in the simulant flowsheet work. Additional work will be needed for the analyses of anions in radiological samples by Analytical Development (AD) and DWPF. The documentation of the development and demonstration of the method fulfills the third requirement in the TTQAP, SRNL-RP-2010-00105, 'Task Technical and Quality Assurance Plan for Glycolic-Formic Acid Flowsheet Development, Definition and Demonstrations Tasks 1-3'.

Best, D.

2010-08-04T23:59:59.000Z

23

Proposed Use of a Constructed Wetland for the Treatment of Metals in the S-04 Outfall of the Defense Waste Processing Facility at the Savannah River Site  

SciTech Connect (OSTI)

The DWPF is part of an integrated waste treatment system at the SRS to treat wastes containing radioactive contaminants. In the early 1980s the DOE recognized that there would be significant safety and cost advantages associated with immobilizing the radioactive waste in a stable solid form. The Defense Waste Processing Facility was designed and constructed to accomplish this task.

Glover, T.

1999-11-23T23:59:59.000Z

24

Basic Data Report -- Defense Waste Processing Facility Sludge Plant, Savannah River Plant 200-S Area  

SciTech Connect (OSTI)

This Basic Data Report for the Defense Waste Processing Facility (DWPF)--Sludge Plant was prepared to supplement the Technical Data Summary. Jointly, the two reports were intended to form the basis for the design and construction of the DWPF. To the extent that conflicting information may appear, the Basic Data Report takes precedence over the Technical Data Summary. It describes project objectives and design requirements. Pertinent data on the geology, hydrology, and climate of the site are included. Functions and requirements of the major structures are described to provide guidance in the design of the facilities. Revision 9 of the Basic Data Report was prepared to eliminate inconsistencies between the Technical Data Summary, Basic Data Report and Scopes of Work which were used to prepare the September, 1982 updated CAB. Concurrently, pertinent data (material balance, curie balance, etc.) have also been placed in the Basic Data Report. It is intended that these balances be used as a basis for the continuing design of the DWPF even though minor revisions may be made in these balances in future revisions to the Technical Data Summary.

Amerine, D.B.

1982-09-01T23:59:59.000Z

25

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

SciTech Connect (OSTI)

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

CHANG, ROBERT

2006-02-02T23:59:59.000Z

26

SUMMARY OF FY11 SULFATE RETENTION STUDIES FOR DEFENSE WASTE PROCESSING FACILITY GLASS  

SciTech Connect (OSTI)

This report describes the results of studies related to the incorporation of sulfate in high level waste (HLW) borosilicate glass produced at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). A group of simulated HLW glasses produced for earlier sulfate retention studies was selected for full chemical composition measurements to determine whether there is any clear link between composition and sulfate retention over the compositional region evaluated. In addition, the viscosity of several glasses was measured to support future efforts in modeling sulfate solubility as a function of predicted viscosity. The intent of these studies was to develop a better understanding of sulfate retention in borosilicate HLW glass to allow for higher loadings of sulfate containing waste. Based on the results of these and other studies, the ability to improve sulfate solubility in DWPF borosilicate glasses lies in reducing the connectivity of the glass network structure. This can be achieved, as an example, by increasing the concentration of alkali species in the glass. However, this must be balanced with other effects of reduced network connectivity, such as reduced viscosity, potentially lower chemical durability, and in the case of higher sodium and aluminum concentrations, the propensity for nepheline crystallization. Future DWPF processing is likely to target higher waste loadings and higher sludge sodium concentrations, meaning that alkali concentrations in the glass will already be relatively high. It is therefore unlikely that there will be the ability to target significantly higher total alkali concentrations in the glass solely to support increased sulfate solubility without the increased alkali concentration causing failure of other Product Composition Control System (PCCS) constraints, such as low viscosity and durability. No individual components were found to provide a significant improvement in sulfate retention (i.e., an increase of the magnitude necessary to have a dramatic impact on blending, washing, or waste loading strategies for DWPF) for the glasses studied here. In general, the concentrations of those species that significantly improve sulfate solubility in a borosilicate glass must be added in relatively large concentrations (e.g., 13 to 38 wt % or more of the frit) in order to have a substantial impact. For DWPF, these concentrations would constitute too large of a portion of the frit to be practical. Therefore, it is unlikely that specific additives may be introduced into the DWPF glass via the frit to significantly improve sulfate solubility. The results presented here continue to show that sulfate solubility or retention is a function of individual glass compositions, rather than a property of a broad glass composition region. It would therefore be inappropriate to set a single sulfate concentration limit for a range of DWPF glass compositions. Sulfate concentration limits should continue to be identified and implemented for each sludge batch. The current PCCS limit is 0.4 wt % SO{sub 4}{sup 2-} in glass, although frit development efforts have led to an increased limit of 0.6 wt % for recent sludge batches. Slightly higher limits (perhaps 0.7-0.8 wt %) may be possible for future sludge batches. An opportunity for allowing a higher sulfate concentration limit at DWPF may lay lie in improving the laboratory experiments used to set this limit. That is, there are several differences between the crucible-scale testing currently used to define a limit for DWPF operation and the actual conditions within the DWPF melter. In particular, no allowance is currently made for sulfur partitioning (volatility versus retention) during melter processing as the sulfate limit is set for a specific sludge batch. A better understanding of the partitioning of sulfur in a bubbled melter operating with a cold cap as well as the impacts of sulfur on the off-gas system may allow a higher sulfate concentration limit to be established for the melter feed. This approach would have to be taken carefully to ensure that a

Fox, K.; Edwards, T.

2012-05-08T23:59:59.000Z

27

Elimination Of Catalytic Hydrogen Generation In Defense Waste Processing Facility Slurries  

SciTech Connect (OSTI)

Based on lab-scale simulations of Defense Waste Processing Facility (DWPF) slurry chemistry, the addition of sodium nitrite and sodium hydroxide to waste slurries at concentrations sufficient to take the aqueous phase into the alkaline region (pH > 7) with approximately 500 mg nitrite ion/kg slurry (assuming <25 wt% total solids, or equivalently 2,000 mg nitrite/kg total solids) is sufficient to effectively deactivate the noble metal catalysts at temperatures between room temperature and boiling. This is a potential strategy for eliminating catalytic hydrogen generation from the list of concerns for sludge carried over into the DWPF Slurry Mix Evaporator Condensate Tank (SMECT) or Recycle Collection Tank (RCT). These conclusions are drawn in large part from the various phases of the DWPF catalytic hydrogen generation program conducted between 2005 and 2009. The findings could apply to various situations, including a solids carry-over from either the Sludge Receipt and Adjustment Tank (SRAT) or Slurry Mix Evaporator (SME) into the SMECT with subsequent transfer to the RCT, as well as a spill of formic acid into the sump system and transfer into an RCT that already contains sludge solids. There are other potential mitigating factors for the SMECT and RCT, since these vessels are typically operated at temperatures close to the minimum temperatures that catalytic hydrogen has been observed to occur in either the SRAT or SME (pure slurry case), and these vessels are also likely to be considerably more dilute in both noble metals and formate ion (the two essential components to catalytic hydrogen generation) than the two primary process vessels. Rhodium certainly, and ruthenium likely, are present as metal-ligand complexes that are favored under certain concentrations of the surrounding species. Therefore, in the SMECT or RCT, where a small volume of SRAT or SME material would be significantly diluted, conditions would be less optimal for forming or sustaining the catalytic ligand species. Such conditions are likely to adversely impact the ability of the transferred mass to produce hydrogen at the same rate (per unit mass SRAT or SME slurry) as in the SRAT or SME vessels.

Koopman, D. C.

2013-01-22T23:59:59.000Z

28

EVALUATION OF A TURBIDITY METER FOR USE AT THE DEFENSE WASTE PROCESSING FACILITY  

SciTech Connect (OSTI)

Savannah River Remediation’s (SRR’s) Defense Waste Processing Facility (DWPF) Laboratory currently tests for sludge carry-over into the Recycle Collection Tank (RCT) by evaluating the iron concentration in the Slurry Mix Evaporator Condensate Tank (SMECT) and relating this iron concentration to the amount of sludge solids present. A new method was proposed for detecting the amount of sludge in the SMECT that involves the use of an Optek turbidity sensor. Waste Services Laboratory (WSL) personnel conducted testing on two of these units following a test plan developed by Waste Solidification Engineering (WSE). Both Optek units (SN64217 and SN65164) use sensor model AF16-N and signal converter model series C4000. The sensor body of each unit was modified to hold a standard DWPF 12 cc sample vial, also known as a “peanut” vial. The purpose of this testing was to evaluate the use of this model of turbidity sensor, or meter, to provide a measurement of the sludge solids present in the SMECT based upon samples from that tank. During discussions of the results from this study by WSE, WSL, and Savannah River National Laboratory (SRNL) personnel, an upper limit on the acceptable level of solids in SMECT samples was set at 0.14 weight percent (wt%). A “go/no-go” decision criterion was to be developed for the critical turbidity response, which is expressed in concentration units (CUs), for each Optek unit based upon the 0.14 wt% solids value. An acceptable or a “go” decision for the SMECT should reflect the situation that there is an identified risk (e.g. 5%) for a CU response from the Optek unit to be less than the critical CU value when the solids content of the SMECT is actually 0.14 wt% or greater, while a “no-go” determination (i.e., an Optek CU response above the critical CU value, a conservative decision relative to risk) would lead to additional evaluations of the SMECT to better quantify the possible solids content of the tank. Subsequent to the issuance of the initial version of this report but under the scope of the original request for technical assistance, WSE asked for this report to be revised to include the “go/no-go” CU value corresponding to 0.28 wt% solids. It was this request that led to the preparation of Revision 1 of the report. The results for the 0.28 wt% solids value were developed following the same approach as that utilized for the 0.14 wt% solids value. A sludge simulant was used to develop standards for testing both Optek units and to determine the viability of a “go/no-go” CU response for each of the units. Statistical methods were used by SRNL to develop the critical CU value for the “go/no-go” decision for these standards for each Optek unit. Since only one sludge simulant was available for this testing, the sensitivity of these results to other simulants and to actual sludge material is not known. However, limited testing with samples from the actual DWPF process (both SRAT product samples and SMECT samples) demonstrated that the use of the “go/no-go” criteria developed from the sludge simulant testing was conservative for these samples taken from the sludge batch, Sludge Batch 7b, being processed at the time of this testing. While both of the Optek units performed very reliably during this testing, there were statistically significant differences (although small on a practical scale) between the two units. Thus, testing should be conducted on any new unit of this Optek model to qualify it before it is used to support the DWPF operation.

Mahannah, R.; Edwards, T.

2013-06-04T23:59:59.000Z

29

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

SciTech Connect (OSTI)

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

Brown, C M

1980-07-01T23:59:59.000Z

30

Qualification of the Nippon Instrumentation for use in Measuring Mercury at the Defense Waste Processing Facility  

SciTech Connect (OSTI)

The Nippon Mercury/RA-3000 system installed in 221-S M-14 has been qualified for use. The qualification was a side-by-side comparison of the Nippon Mercury/RA-3000 system with the currently used Bacharach Mercury Analyzer. The side-by-side testing included standards for instrument calibration verifications, spiked samples and unspiked samples. The standards were traceable back to the National Institute of Standards and Technology (NIST). The side-by-side work included the analysis of Sludge Receipt and Adjustment Tank (SRAT) Receipt, SRAT Product, and Slurry Mix Evaporator (SME) samples. With the qualification of the Nippon Mercury/RA-3000 system in M-14, the DWPF lab will be able to perform a head to head comparison of a second Nippon Mercury/RA-3000 system once the system is installed. The Defense Waste Processing Facility (DWPF) analyzes receipt and product samples from the Sludge Receipt and Adjustment Tank (SRAT) to determine the mercury (Hg) concentration in the sludge slurry. The SRAT receipt is typically sampled and analyzed for the first ten SRAT batches of a new sludge batch to obtain an average Hg concentration. This average Hg concentration is then used to determine the amount of steam stripping required during the concentration/reflux step of the SRAT cycle to achieve a less than 0.6 wt% Hg in the SRAT product solids. After processing is complete, the SRAT product is sampled and analyzed for mercury to ensure that the mercury concentration does not exceed the 0.45 wt% limit in the Slurry Mix Evaporator (SME). The DWPF Laboratory utilizes Bacharach Analyzers to support these Hg analyses at this facility. These analyzers are more than 10 years old, and they are no longer supported by the manufacturer. Due to these difficulties, the Bacharach Analyzers are to be replaced by new Nippon Mercury/RA-3000 systems. DWPF issued a Technical Task Request (TTR) for the Savannah River National Laboratory (SRNL) to assist in the qualification of the new systems. SRNL prepared a task technical and quality assurance (TT&QA) plan that outlined the activities that are necessary and sufficient to meet the objectives of the TTR. In addition, TT&QA plan also included a test plan that provided guidance to the DWPF Lab in collecting the data needed to qualify the new Nippon Mercury/RA-3000 systems.

Edwards, T.; Mahannah, R.

2011-07-05T23:59:59.000Z

31

Qualification of the First ICS-3000 ION Chromatograph for use at the Defense Waste Processing Facility  

SciTech Connect (OSTI)

The ICS-3000 Ion Chromatography (IC) system installed in 221-S M-13 has been qualified for use. The qualification was a head to head comparison of the ICS-3000 with the currently used DX-500 IC system. The crosscheck work included standards for instrument calibration and calibration verifications and standards for individual anion analysis, where the standards were traceable back to the National Institute of Standards and Technology (NIST). In addition the crosscheck work included the analysis of simulated Sludge Receipt Adjustment Tank (SRAT) Receipt, SRAT Product, and Slurry Mix Evaporator (SME) samples, along with radioactive Sludge Batch 5 material from the SRAT and SME tanks. Based upon the successful qualification of the ICS-3000 in M-13, it is recommended that this task proceed in developing the data to qualify, by a head to head comparison of the two ICS-3000 instruments, a second ICS-3000 to be installed in M-14. The Defense Waste Processing Facility (DWPF) requires the analysis of specific anions at various stages of its processing of high level waste (HLW). The anions of interest to the DWPF are fluoride, formate, chloride, nitrite, nitrate, sulfate, oxalate, and phosphate. The anion analysis is used to evaluate process chemistry including formic acid/nitric acid additions to establish optimum conditions for mercury stripping, reduction-oxidation (REDOX) chemistry for the melter, nitrite destruction, organic acid constituents, etc. The DWPF Laboratory (Lab) has been using Dionex DX-500 ion chromatography (IC) systems since 1998. The vendor informed DWPF in 2006 that the instruments would no longer be supported by service contracts after 2008. DWPF purchased three new ICS-3000 systems in September of 2006. The ICS-3000 instruments are (a) designed to be more stable using an eluent generator to make eluent, (b) require virtually no daily chemical handling by the analysts, (c) require less line breaks in the hood, and (d) generally require less maintenance due to the pump configuration only using water versus the current system where the pump uses various hydroxide concentrations. The ICS-3000 instruments also allow the DWPF to maintain current service contracts, which support routine preventive maintenance and emergency support for larger problems such as component failure. One of the three new systems was set up in the DWPF Lab trailers in January of 2007 to be used for the development of methods and procedures. This system will continue to be used for training, new method development and potential improvements to current methods. The qualification of the other two ICS-3000 instruments is to be a phased effort. This effort is to be supported by the Applied Computational Engineering and Statistical (ACES) group of the Savannah River National Laboratory (SRNL) as authorized by the Technical Task Request (TTR) and as directed by the corresponding Task Technical and Quality Assurance (TT&QA) plan. The installation of the first 'rad' system into the M-13 Lab module required modifications to both the Lab module and to the radiohood. The installation was completed in July 2008. The testing of this system was conducted as directed by the TT&QA plan. The purpose of this technical report is to provide a review of the data generated by these tests that will lead to the recommendation for the qualification of the M-13 ICS-3000 instrument. With the successful qualification of this first ICS-3000, plans will be developed for the installation of the second 'rad' system in the M-14 Lab module later in fiscal year 2009. When the second 'rad' ICS-3000 system is installed, the DX-500 systems will be removed and retired from service.

Edwards, T; Mahannah, R.

2011-07-05T23:59:59.000Z

32

VERIFICATION OF THE DEFENSE WASTE PROCESSING FACILITY'S (DWPF) PROCESS DIGESTION METHOD FOR THE SLUDGE BATCH 7A QUALIFICATION SAMPLE  

SciTech Connect (OSTI)

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs confirmation of the applicability of the digestion method to be used by the DWPF lab for elemental analysis of Sludge Receipt and Adjustment Tank (SRAT) receipt samples and SRAT product process control samples. DWPF SRAT samples are typically dissolved using a room temperature HF-HNO{sub 3} acid dissolution (i.e., DWPF Cold Chem Method, see DWPF Procedure SW4-15.201) and then analyzed by inductively coupled plasma - atomic emission spectroscopy (ICP-AES). This report contains the results and comparison of data generated from performing the Aqua Regia (AR), Sodium peroxide/Hydroxide Fusion (PF) and DWPF Cold Chem (CC) method digestions of Sludge Batch 7a (SB7a) SRAT Receipt and SB7a SRAT Product samples. The SB7a SRAT Receipt and SB7a SRAT Product samples were prepared in the SRNL Shielded Cells, and the SRAT Receipt material is representative of the sludge that constituates the SB7a Batch or qualification composition. This is the sludge in Tank 51 that is to be transferred into Tank 40, which will contain the heel of Sludge Batch 6 (SB6), to form the Sb7a Blend composition.

Click, D.; Edwards, T.; Jones, M.; Wiedenman, B.

2011-03-14T23:59:59.000Z

33

CHARACTERIZATION OF A PRECIPITATE REACTOR FEED TANK (PRFT) SAMPLE FROM THE DEFENSE WASTE PROCESSING FACILITY (DWPF)  

SciTech Connect (OSTI)

A sample of from the Defense Waste Processing Facility (DWPF) Precipitate Reactor Feed Tank (PRFT) was pulled and sent to the Savannah River National Laboratory (SRNL) in June of 2013. The PRFT in DWPF receives Actinide Removal Process (ARP)/ Monosodium Titanate (MST) material from the 512-S Facility via the 511-S Facility. This 2.2 L sample was to be used in small-scale DWPF chemical process cell testing in the Shielded Cells Facility of SRNL. A 1L sub-sample portion was characterized to determine the physical properties such as weight percent solids, density, particle size distribution and crystalline phase identification. Further chemical analysis of the PRFT filtrate and dissolved slurry included metals and anions as well as carbon and base analysis. This technical report describes the characterization and analysis of the PRFT sample from DWPF. At SRNL, the 2.2 L PRFT sample was composited from eleven separate samples received from DWPF. The visible solids were observed to be relatively quick settling which allowed for the rinsing of the original shipping vials with PRFT supernate on the same day as compositing. Most analyses were performed in triplicate except for particle size distribution (PSD), X-ray diffraction (XRD), Scanning Electron Microscopy (SEM) and thermogravimetric analysis (TGA). PRFT slurry samples were dissolved using a mixed HNO3/HF acid for subsequent Inductively Coupled Plasma Atomic Emission Spectroscopy (ICPAES) and Inductively Coupled Plasma Mass Spectroscopy (ICP-MS) analyses performed by SRNL Analytical Development (AD). Per the task request for this work, analysis of the PRFT slurry and filtrate for metals, anions, carbon and base were primarily performed to support the planned chemical process cell testing and to provide additional component concentrations in addition to the limited data available from DWPF. Analysis of the insoluble solids portion of the PRFT slurry was aimed at detailed characterization of these solids (TGA, PSD, XRD and SEM) in support of the Salt IPT chemistry team. The overall conclusions from analyses performed in this study are that the PRFT slurry consists of 0.61 Wt.% insoluble MST solids suspended in a 0.77 M [Na+] caustic solution containing various anions such as nitrate, nitrite, sulfate, carbonate and oxalate. The corresponding measured sulfur level in the PRFT slurry, a critical element for determining how much of the PRFT slurry gets blended into the SRAT, is 0.437 Wt.% TS. The PRFT slurry does not contain insoluble oxalates nor significant quantities of high activity sludge solids. The lack of sludge solids has been alluded to by the Salt IPT chemistry team in citing that the mixing pump has been removed from Tank 49H, the feed tank to ARP-MCU, thus allowing the sludge solids to settle out. ? The PRFT aqueous slurry from DWPF was found to contain 5.96 Wt.% total dried solids. Of these total dried solids, relatively low levels of insoluble solids (0.61 Wt.%) were measured. The densities of both the filtrate and slurry were 1.05 g/mL. ? Particle size distribution of the PRFT solids in filtered caustic simulant and XRD analysis of washed/dried PRFT solids indicate that the PRFT slurry contains a bimodal distribution of particles in the range of 1 and 6 ?m and that the particles contain sodium titanium oxide hydroxide Na2Ti2O4(OH)2 crystalline material as determined by XRD. These data are in excellent agreement with similar data obtained from laboratory sampling of vendor supplied MST. Scanning Electron Microscopy (SEM) combined with Energy Dispersive X-ray Spectroscopy (EDS) analysis of washed/dried PRFT solids shows the particles to be like previous MST analyses consisting of irregular shaped micron-sized solids consisting primarily of Na and Ti. ? Thermogravimetric analysis of the washed and unwashed PRFT solids shows that the washed solids are very similar to MST solids. The TGA mass loss signal for the unwashed solids shows similar features to TGA performed on cellulose nitrate filter paper indicating significant presence of the deteriorated filter

Crawford, C.; Bannochie, C.

2014-05-12T23:59:59.000Z

34

EIS-0074: Long-Term Management of Defense High-Level Radioactive Wastes Idaho Chemical Processing Plant, Idaho National Engineering Lab, Idaho  

Broader source: Energy.gov [DOE]

The U.S. Department of Energy prepared this statement to analyze the environmental implications of the proposed selection of a strategy for long- term management of the high- level radioactive wastes generated as part of the national defense effort at the Department's Idaho Chemical Processing Plant a t the Idaho National Engineering Laboratory.

35

Electro-Mechanical Manipulator for Use in the Remote Equipment Decontamination Cell at the Defense Waste Processing Facility, Savannah River Site - 12454  

SciTech Connect (OSTI)

One of the legacies of the cold war is millions of liters of radioactive waste. One of the locations where this waste is stored is at the Savannah River Site (SRS) in South Carolina. A major effort to clean up this waste is on-going at the defense waste processing facility (DWPF) at SRS. A piece of this effort is decontamination of the equipment used in the DWPF to process the waste. The remote equipment decontamination cell (REDC) in the DWPF uses electro-mechanical manipulators (EMM) arms manufactured and supplied by PaR Systems to decontaminate DWPF process equipment. The decontamination fluid creates a highly corrosive environment. After 25 years of operational use the original EMM arms are aging and need replacement. To support continued operation of the DWPF, two direct replacement EMM arms were delivered to the REDC in the summer of 2011. (authors)

Lambrecht, Bill; Dixon, Joe [Par Systems, Shoreview, Minnesota, 55126 (United States); Neuville, John R. [Savannah River Remediation, Savannah River Site, Aiken, South Carolina, 29808 (United States)

2012-07-01T23:59:59.000Z

36

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

SciTech Connect (OSTI)

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

Not Available

1982-09-01T23:59:59.000Z

37

Analyses by the Defense Waste Processing Facility Laboratory of Thorium Glasses from the Sludge Batch 6 Variability Study  

SciTech Connect (OSTI)

The Savannah River Remediation (SRR) Defense Waste Processing Facility (DWPF) is currently processing Sludge Batch 6 (SB6) with Frit 418. At times during the processing of this glass system, thorium is expected to be at concentrations in the final wasteform that make it a reportable element for the first time since startup of radioactive operations at the DWPF. The Savannah River National Laboratory (SRNL) supported the qualification of the processing of this glass system at the DWPF. A recommendation from the SRNL studies was the need for the DWPF Laboratory to establish a method to measure thorium by Inductively Coupled Plasma - Atomic Emission Spectroscopy (ICPAES). This recommendation led to the set of thorium-bearing glasses from the SB6 Variability Study (VS) being submitted to the DWPF Laboratory for chemical composition measurement. The measurements were conducted by the DWPF Laboratory using the sodium peroxide fusion preparation method routinely employed for analysis of samples from the Slurry Mix Evaporator (SME). These measurements are presented and reviewed in this report. The review indicates that the measurements provided by the DWPF Laboratory are comparable to those provided by Analytical Development's laboratory at SRNL for these same glasses. As a result, the authors of this report recommend that the DWPF Laboratory begin using its routine peroxide fusion dissolution method for the measurement of thorium in SME samples of SB6. The purpose of this technical report is to present the measurements generated by the DWPF Laboratory for the SB6 VS glasses and to compare the measurements to the targeted compositions for these VS glasses as well as to SRNL's measurements (both sets, targeted and measured, of compositional values were reported by SRNL in [2]). The goal of these comparisons is to provide information that will lead to the qualification of peroxide fusion dissolution as a method for the measurement by the DWPF Laboratory of thorium in SME glass samples.

Edwards, T.; Click, D.; Feller, M.

2011-02-28T23:59:59.000Z

38

Defense waste transportation: cost and logistics studies  

SciTech Connect (OSTI)

Transportation of nuclear wastes from defense programs is expected to significantly increase in the 1980s and 1990s as permanent waste disposal facilities come into operation. This report uses models of the defense waste transportation system to quantify potential transportation requirements for treated and untreated contact-handled transuranic (CH-TRU) wastes and high-level defense wastes (HLDW). Alternative waste management strategies in repository siting, waste retrieval and treatment, treatment facility siting, waste packaging and transportation system configurations were examined to determine their effect on transportation cost and hardware requirements. All cost estimates used 1980 costs. No adjustments were made for future changes in these costs relative to inflation. All costs are reported in 1980 dollars. If a single repository is used for defense wastes, transportation costs for CH-TRU waste currently in surface storage and similar wastes expected to be generated by the year 2000 were estimated to be 109 million dollars. Recovery and transport of the larger buried volumes of CH-TRU waste will increase CH-TRU waste transportation costs by a factor of 70. Emphasis of truck transportation and siting of multiple repositories would reduce CH-TRU transportation costs. Transportation of HLDW to repositories for 25 years beginning in 1997 is estimated to cost $229 M in 1980 costs and dollars. HLDW transportation costs could either increase or decrease with the selection of a final canister configuration. HLDW transportation costs are reduced when multiple repositories exist and emphasis is placed on truck transport.

Andrews, W.B.; Cole, B.M.; Engel, R.L.; Oylear, J.M.

1982-08-01T23:59:59.000Z

39

MEASUREMENT AND CALCULATION OF RADIONUCLIDE ACTIVITIES IN SAVANNAH RIVER SITE HIGH LEVEL WASTE SLUDGE FOR ACCEPTANCE OF DEFENSE WASTE PROCESSING FACILITY GLASS IN A FEDERAL REPOSITORY  

SciTech Connect (OSTI)

This paper describes the results of the analyses of High Level Waste (HLW) sludge slurry samples and of the calculations necessary to decay the radionuclides to meet the reporting requirement in the Waste Acceptance Product Specifications (WAPS) [1]. The concentrations of 45 radionuclides were measured. The results of these analyses provide input for radioactive decay calculations used to project the radionuclide inventory at the specified index years, 2015 and 3115. This information is necessary to complete the Production Records at Savannah River Site's Defense Waste Processing Facility (DWPF) so that the final glass product resulting from Macrobatch 5 (MB5) can eventually be submitted to a Federal Repository. Five of the necessary input radionuclides for the decay calculations could not be measured directly due to their low concentrations and/or analytical interferences. These isotopes are Nb-93m, Pd-107, Cd-113m, Cs-135, and Cm-248. Methods for calculating these species from concentrations of appropriate other radionuclides will be discussed. Also the average age of the MB5 HLW had to be calculated from decay of Sr-90 in order to predict the initial concentration of Nb-93m. As a result of the measurements and calculations, thirty-one WAPS reportable radioactive isotopes were identified for MB5. The total activity of MB5 sludge solids will decrease from 1.6E+04 {micro}Ci (1 {micro}Ci = 3.7E+04 Bq) per gram of total solids in 2008 to 2.3E+01 {micro}Ci per gram of total solids in 3115, a decrease of approximately 700 fold. Finally, evidence will be given for the low observed concentrations of the radionuclides Tc-99, I-129, and Sm-151 in the HLW sludges. These radionuclides were reduced in the MB5 sludge slurry to a fraction of their expected production levels due to SRS processing conditions.

Bannochie, C; David Diprete, D; Ned Bibler, N

2008-12-31T23:59:59.000Z

40

Bubblers Speed Nuclear Waste Processing at SRS  

SciTech Connect (OSTI)

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

None

2010-11-14T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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

Bubblers Speed Nuclear Waste Processing at SRS  

ScienceCinema (OSTI)

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

None

2014-08-06T23:59:59.000Z

42

Defense waste vitrification studies during FY-1981. Summary report  

SciTech Connect (OSTI)

Both simulated alkaline defense wastes and simulated acidic defense wastes (formed by treating alkaline waste with formic acid) were successfully vitrified in direct liquid-fed melter experiments. The vitrification process was improved while using the formate-treated waste. Leach resistance was essentially the same. Off-gas entrainment was the primary mechanism for material exiting the melter. When formate waste was vitrified, the flow behavior of the off gas from the melter changed dramatically from an erratic surging behavior to a more quiet, even flow. Hydrogen and CO were detectable while processing formate feed; however, levels exceeding the flamability limits in air were never approached. Two types of melter operation were tested during the year, one involving boost power. Several boosting methods located within the melter plenum were tested. When lid heating was being used, water spray cooling in the off gas was required. Countercurrent spray cooling was more effective than cocurrent spray cooling. Materials of construction for the off-gas system were examined. Inconel-690 is preferred in the plenum area. Inspection of the pilot-scale melter found that corrosion of the K-3 refractory and Inconel-690 electrodes was minimal. An overheating incident occurred with the LFCM in which glass temperatures up to 1480/sup 0/C were experienced. Lab-scale vitrification tests to study mercury behavior were also completed this year. 53 figures, 63 tables.

Bjorklund, W.J. (comp.)

1982-09-01T23:59:59.000Z

43

EVALUATION OF THE IMPACT OF THE DEFENSE WASTE PROCESSING FACILITY (DWPF) LABORATORY GERMANIUM OXIDE USE ON RECYCLE TRANSFERS TO THE H-TANK FARM  

SciTech Connect (OSTI)

When processing High Level Waste (HLW) glass, the Defense Waste Processing Facility (DWPF) cannot wait until the melt or waste glass has been made to assess its acceptability, since by then no further changes to the glass composition and acceptability are possible. Therefore, the acceptability decision is made on the upstream feed stream, rather than on the downstream melt or glass product. This strategy is known as 'feed forward statistical process control.' The DWPF depends on chemical analysis of the feed streams from the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mix Evaporator (SME) where the frit plus adjusted sludge from the SRAT are mixed. The SME is the last vessel in which any chemical adjustments or frit additions can be made. Once the analyses of the SME product are deemed acceptable, the SME product is transferred to the Melter Feed Tank (MFT) and onto the melter. The SRAT and SME analyses have been analyzed by the DWPF laboratory using a 'Cold Chemical' method but this dissolution did not adequately dissolve all the elemental components. A new dissolution method which fuses the SRAT or SME product with cesium nitrate (CsNO{sub 3}), germanium (IV) oxide (GeO{sub 2}) and cesium carbonate (Cs{sub 2}CO{sub 3}) into a cesium germanate glass at 1050 C in platinum crucibles has been developed. Once the germanium glass is formed in that fusion, it is readily dissolved by concentrated nitric acid (about 1M) to solubilize all the elements in the SRAT and/or SME product for elemental analysis. When the chemical analyses are completed the acidic cesium-germanate solution is transferred from the DWPF analytic laboratory to the Recycle Collection Tank (RCT) where the pH is increased to {approx}12 M to be released back to the tank farm and the 2H evaporator. Therefore, about 2.5 kg/yr of GeO{sub 2}/year will be diluted into 1.4 million gallons of recycle. This 2.5 kg/yr of GeO{sub 2} may increase to 4 kg/yr when improvements are implemented to attain an annual canister production goal of 400 canisters. Since no Waste Acceptance Criteria (WAC) exists for germanium in the Tank Farm, the Effluent Treatment Project, or the Saltstone Production Facility, DWPF has requested an evaluation of the fate of the germanium in the caustic environment of the RCT, the 2H evaporator, and the tank farm. This report evaluates the effect of the addition of germanium to the tank farm based on: (1) the large dilution of Ge in the RCT and tank farm; (2) the solubility of germanium in caustic solutions (pH 12-13); (3) the potential of germanium to precipitate as germanium sodalites in the 2H Evaporator; and (4) the potential of germanium compounds to precipitate in the evaporator feed tank. This study concludes that the impacts of transferring up to 4 kg/yr germanium to the RCT (and subsequently the 2H evaporator feed tank and the 2H evaporator) results in <2 ppm per year (1.834 mg/L) which is the maximum instantaneous concentration expected from DWPF. This concentration is insignificant as most sodium germanates are soluble at the high pH of the feed tank and evaporator solutions. Even if sodium aluminosilicates form in the 2H evaporator, the Ge will likely substitute for some small amount of the Si in these structures and will be insignificant. It is recommended that the DWPF continue with their strategy to add germanium as a laboratory chemical to Attachment 8.2 of the DWPF Waste Compliance Plan (WCP).

Jantzen, C.; Laurinat, J.

2011-08-15T23:59:59.000Z

44

Waste processing air cleaning  

SciTech Connect (OSTI)

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

Kriskovich, J.R.

1998-07-27T23:59:59.000Z

45

Salt Waste Processing Initiatives  

Office of Environmental Management (EM)

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

46

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

SciTech Connect (OSTI)

The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available.

Not Available

1983-06-01T23:59:59.000Z

47

MODELING THE IMPACT OF ELEVATED MERCURY IN DEFENSE WASTE PROCESSING FACILITY MELTER FEED ON THE MELTER OFF-GAS SYSTEM - PRELIMINARY REPORT  

SciTech Connect (OSTI)

The Defense Waste Processing Facility (DWPF) is currently evaluating an alternative Chemical Process Cell (CPC) flowsheet to increase throughput. It includes removal of the steam-stripping step, which would significantly reduce the CPC processing time and lessen the sampling needs. However, its downside would be to send 100% of the mercury that come in with the sludge straight to the melter. For example, the new mercury content in the Sludge Batch 5 (SB5) melter feed is projected to be 25 times higher than that in the SB4 with nominal steam stripping of mercury. This task was initiated to study the impact of the worst-case scenario of zero-mercury-removal in the CPC on the DWPF melter off-gas system. It is stressed that this study is intended to be scoping in nature, so the results presented in this report are preliminary. In order to study the impact of elevated mercury levels in the feed, it is necessary to be able to predict how mercury would speciate in the melter exhaust under varying melter operating conditions. A homogeneous gas-phase oxidation model of mercury by chloride was developed to do just that. The model contains two critical parameters pertaining to the partitioning of chloride among HCl, Cl, Cl{sub 2}, and chloride salts in the melter vapor space. The values for these parameters were determined at two different melter vapor space temperatures by matching the calculated molar ratio of HgCl (or Hg{sub 2}Cl{sub 2}) to HgCl{sub 2} with those measured during the Experimental-Scale Ceramic Melter (ESCM) tests run at the Pacific Northwest National Laboratory (PNNL). The calibrated model was then applied to the SB5 simulant used in the earlier flowsheet study with an assumed mercury stripping efficiency of zero; the molar ratio of Cl-to-Hg in the resulting melter feed was only 0.4, compared to 12 for the ESCM feeds. The results of the model run at the indicated melter vapor space temperature of 650 C (TI4085D) showed that due to excessive shortage of chloride, only 6% of the mercury fed is expected to get oxidized, mostly as HgCl, while the remaining mercury would exist either as elemental mercury vapor (90%) or HgO (4%). Noting that the measured chloride level in the SB5 qualification sample was an order of magnitude lower than that used in the SB5 simulant, the degree of chloride shortage will be even greater. As a result, the projected level of HgCl in the actual SB5 melter exhaust will be even lower than 6% of the total mercury fed, while that of elemental mercury is likely to be greater than 90%. The homogeneous oxidation of mercury in the off-gas was deemed to be of primary importance based on the postulation that mercury and other volatile salts form submicron sized aerosols upon condensation and thus remain largely in the gas stream downstream of the quencher where they can deposit in the off-gas lines, Steam-Atomized Scrubbers (SAS), and High-Efficiency Mist Eliminator (HEME). Formation of these submicron semi-volatile salts in the condensate liquid is considered to be unlikely, so the liquid phase reactions were considered to be less important. However, subsequent oxidation of mercury in the liquid phase in the off-gas system was examined in a simplified model of the off-gas condensate. It was found that the condensate chemistry was consistent with further oxidation of elemental mercury to Hg{sub 2}Cl{sub 2} and conversion of HgO to chlorides. The results were consistent with the available experimental data. It should also be noted that the model predictions presented in this report do not include any physically entrained solids, which typically account for much of the off-gas carryover on a mass basis. The high elemental mercury vapor content predicted at the DWPF Quencher inlet means that physically entrained solids could provide the necessary surface onto which elemental mercury vapor could condense, thereby coating the solids as well as the internal surfaces of the off-gas system with mercury. Clearly, there are many process benefits to be gained by removing the steam-stripping step from the CPC c

Zamecnik, J.; Choi, A.

2009-03-25T23:59:59.000Z

48

MODELING THE IMPACT OF ELEVATED MERCURY IN DEFENSE WASTE PROCESSING FACILITY MELTER FEED ON THE MELTER OFF-GAS SYSTEM-PRELIMINARY REPORT  

SciTech Connect (OSTI)

The Defense Waste Processing Facility (DWPF) is currently evaluating an alternative Chemical Process Cell (CPC) flowsheet to increase throughput. It includes removal of the steam-stripping step, which would significantly reduce the CPC processing time and lessen the sampling needs. However, its downside would be to send 100% of the mercury that comes in with the sludge straight to the melter. For example, the new mercury content in the Sludge Batch 5 (SB5) melter feed is projected to be 25 times higher than that in the SB4 with nominal steam stripping of mercury. This task was initiated to study the impact of the worst-case scenario of zero-mercury-removal in the CPC on the DWPF melter offgas system. It is stressed that this study is intended to be scoping in nature, so the results presented in this report are preliminary. In order to study the impact of elevated mercury levels in the feed, it is necessary to be able to predict how mercury would speciate in the melter exhaust under varying melter operating conditions. A homogeneous gas-phase oxidation model of mercury by chloride was developed to do just that. The model contains two critical parameters pertaining to the partitioning of chloride among HCl, Cl, Cl{sub 2}, and chloride salts in the melter vapor space. The values for these parameters were determined at two different melter vapor space temperatures by matching the calculated molar ratio of HgCl (or Hg{sub 2}Cl{sub 2}) to HgCl{sub 2} with those measured during the Experimental-Scale Ceramic Melter (ESCM) tests run at the Pacific Northwest National Laboratory (PNNL). The calibrated model was then applied to the SB5 simulant used in the earlier flowsheet study with an assumed mercury stripping efficiency of zero; the molar ratio of Cl-to-Hg in the resulting melter feed was only 0.4, compared to 12 for the ESCM feeds. The results of the model run at the indicated melter vapor space temperature of 650 C (TI4085D) showed that due to excessive shortage of chloride, only 6% of the mercury fed is expected to get oxidized, mostly as HgCl, while the remaining mercury would exist either as elemental mercury vapor (90%) or HgO (4%). Noting that the measured chloride level in the SB5 qualification sample was an order of magnitude lower than that used in the SB5 simulant, the degree of chloride shortage will be even greater. As a result, the projected level of HgCl in the actual SB5 melter exhaust will be even lower than 6% of the total mercury fed, while that of elemental mercury is likely to be greater than 90%. The homogeneous oxidation of mercury in the off-gas was deemed to be of primary importance based on the postulation that mercury and other volatile salts form submicron sized aerosols upon condensation and thus remain largely in the gas stream downstream of the quencher where they can deposit in the off-gas lines, Steam-Atomized Scrubbers (SAS), and High-Efficiency Mist Eliminator (HEME). Formation of these submicron semi-volatile salts in the condensate liquid is considered to be unlikely, so the liquid phase reactions were considered to be less important. However, subsequent oxidation of mercury in the liquid phase in the off-gas system was examined in a simplified model of the off-gas condensate. It was found that the condensate chemistry was consistent with further oxidation of elemental mercury to Hg{sub 2}Cl{sub 2} and conversion of HgO to chlorides. The results were consistent with the available experimental data. It should also be noted that the model predictions presented in this report do not include any physically entrained solids, which typically account for much of the off-gas carryover on a mass basis. The high elemental mercury vapor content predicted at the DWPF Quencher inlet means that physically entrained solids could provide the necessary surface onto which elemental mercury vapor could condense, thereby coating the solids as well as the internal surfaces of the off-gas system with mercury. Clearly, there are many process benefits to be gained by removing the steam-stripping step from the CPC c

Zamecnik, J.; Choi, A.

2010-08-18T23:59:59.000Z

49

Long-range master plan for defense transuranic waste management  

SciTech Connect (OSTI)

The Long Range Master Plan for the Defense Transuranic Waste Program (DTWP), or ''Master Plan,'' details current TRU waste management plans and serves as a framework for the DTWP. Not all final decisions concerning activities presented in the Master Plan have been made (e.g., land withdrawal legislation, the WIPP Compliance and Operational Plan and the TRUPACT Certificate of Compliance). It is the goal of the DTWP to end interim storage and achieve permanent disposal of TRU waste. To accomplish this goal, as much TRU waste as possible will be certified to meet the WIPP Acceptance Criteria (WAC). The certified waste will then be disposed of at WIPP. The small quantity of waste which is not practical to certify will be disposed of via alternative methods that require DOE Headquarters approval and shall comply with the National Environmental Policy Act requirements and EPA/State Regulations. The definition of TRU waste is ''without regard to source or form, waste that is contaminated with alpha-emitting transuranium radionuclides with half-lives greater than 20 years and concentrations greater than 100 nanocuries/gram (nCi/g) at the time of assay. Heads of Field Elements can determine that other alpha contaminated wastes, peculiar to a specific site, must be managed as transuranic waste.''

Not Available

1988-12-01T23:59:59.000Z

50

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

51

Defense High-Level Waste Leaching Mechanisms Program. Final report  

SciTech Connect (OSTI)

The Defense High-Level Waste Leaching Mechanisms Program brought six major US laboratories together for three years of cooperative research. The participants reached a consensus that solubility of the leached glass species, particularly solubility in the altered surface layer, is the dominant factor controlling the leaching behavior of defense waste glass in a system in which the flow of leachant is constrained, as it will be in a deep geologic repository. Also, once the surface of waste glass is contacted by ground water, the kinetics of establishing solubility control are relatively rapid. The concentrations of leached species reach saturation, or steady-state concentrations, within a few months to a year at 70 to 90/sup 0/C. Thus, reaction kinetics, which were the main subject of earlier leaching mechanisms studies, are now shown to assume much less importance. The dominance of solubility means that the leach rate is, in fact, directly proportional to ground water flow rate. Doubling the flow rate doubles the effective leach rate. This relationship is expected to obtain in most, if not all, repository situations.

Mendel, J.E. (compiler)

1984-08-01T23:59:59.000Z

52

Terminating Safeguards on Excess Special Nuclear Material: Defense TRU Waste Clean-up and Nonproliferation - 12426  

SciTech Connect (OSTI)

The Department of Energy (DOE) and the National Nuclear Security Administration (NNSA) manages defense nuclear material that has been determined to be excess to programmatic needs and declared waste. When these wastes contain plutonium, they almost always meet the definition of defense transuranic (TRU) waste and are thus eligible for disposal at the Waste Isolation Pilot Plant (WIPP). The DOE operates the WIPP in a manner that physical protections for attractiveness level D or higher special nuclear material (SNM) are not the normal operating condition. Therefore, there is currently a requirement to terminate safeguards before disposal of these wastes at the WIPP. Presented are the processes used to terminate safeguards, lessons learned during the termination process, and how these approaches might be useful for future defense TRU waste needing safeguards termination prior to shipment and disposal at the WIPP. Also described is a new criticality control container, which will increase the amount of fissile material that can be loaded per container, and how it will save significant taxpayer dollars. Retrieval, compliant packaging and shipment of retrievably stored legacy TRU waste has dominated disposal operations at WIPP since it began operations 12 years ago. But because most of this legacy waste has successfully been emplaced in WIPP, the TRU waste clean-up focus is turning to newly-generated TRU materials. A major component will be transuranic SNM, currently managed in safeguards-protected vaults around the weapons complex. As DOE and NNSA continue to consolidate and shrink the weapons complex footprint, it is expected that significant quantities of transuranic SNM will be declared surplus to the nation's needs. Safeguards termination of SNM varies due to the wide range of attractiveness level of the potential material that may be directly discarded as waste. To enhance the efficiency of shipping waste with high TRU fissile content to WIPP, DOE designed an over-pack container, similar to the pipe component, called the criticality control over-pack, which will significantly enhance the efficiency of disposal. Hundreds of shipments of transuranic SNM, suitably packaged to meet WIPP waste acceptance criteria and with safeguards terminated have been successfully emplaced at WIPP (primarily from the Rocky Flats site clean-up) since WIPP opened. DOE expects that thousands more may eventually result from SNM consolidation efforts throughout the weapons complex. (authors)

Hayes, Timothy [Los Alamos National Laboratory, Carlsbad Operations Group (United States); Nelson, Roger [Department Of Energy, Carlsbad Operations Office (United States)

2012-07-01T23:59:59.000Z

53

Feasibility of very deep borehole disposal of US nuclear defense wastes  

E-Print Network [OSTI]

This thesis analyzes the feasibility of emplacing DOE-owned defense nuclear waste from weapons production into a permanent borehole repository drilled ~4 km into granite basement rock. Two canister options were analyzed ...

Dozier, Frances Elizabeth

2011-01-01T23:59:59.000Z

54

Fiscal year 1986 program plan for the Defense Transuranic Waste Program (DTWP)  

SciTech Connect (OSTI)

The Defense TRU Waste Program (DTWP) is the focal point for the Department of Energy is national planning, integration, and technical development for TRU waste management. The scope of this program extends from the point of TRU waste generation through delivery to a permanent repository. The TRU program maintains a close interface with repository development to ensure program compatibility and coordination. The defense TRU program does not directly address commercial activities that generate TRU waste. Instead, it is concerned with providing alternatives to manage existing and future defense TRU wastes. The FY 86 Program Plan is consistent with the Defense TRU Waste Program goals and objectives stated in the Defense Transuranic Waste Program Strategy Document, January 1984. The roles of participants, the responsibilities and authorities for Research Development (R D), the organizational interfaces and communication channels for R D and the establishment of procedures for planning, reporting, and budgeting of all R D activities meet requirements tated in the Technical Management Plan for the Transuranic Waste Management Program. The Program Plan is revised as needed. Detailed budget planning (i.e., programmatic funding and capital equipment) is presented for FY 86; outyear budget projections are presented for future years.

Not Available

1985-11-01T23:59:59.000Z

55

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

56

Hydrothermal Processing of Wet Wastes  

Broader source: Energy.gov [DOE]

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

57

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

SciTech Connect (OSTI)

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

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

1982-09-01T23:59:59.000Z

58

EIS-0063: Waste Management Operations, Double-Shell Tanks for Defense High Level Radioactive Waste Storage, Hanford Site, Richland, Washington  

Broader source: Energy.gov [DOE]

The U.S. Department of Energy developed this statement to evaluate the existing tank design and consider additional specific design and safety feature alternatives for the thirteen tanks being constructed for storage of defense high-level radioactive liquid waste at the Hanford Site in Richland, Washington. This statement supplements ERDA-1538, "Final Environmental Statement on Waste Management Operation."

59

Potential radiological impacts of upper-bound operational accidents during proposed waste disposal alternatives for Hanford defense waste  

SciTech Connect (OSTI)

The Geologic Disposal Alternative, the In-Place Stabilization and Disposal Alternative, and the Reference Disposal Alternative are being evaluated for disposal of Hanford defense high-level, transuranic, and tank wastes. Environmental impacts associated with disposal of these wastes according to the alternatives listed above include potential doses to the downwind population from operation during the application of the handling and processing techniques comprising each disposal alternative. Scenarios for operational accident and abnormal operational events are postulated, on the basis of the currently available information, for the application of the techniques employed for each waste class for each disposal alternative. From these scenarios, an upper-bound airborne release of radioactive material was postulated for each waste class and disposal alternative. Potential downwind radiologic impacts were calculated from these upper-bound events. In all three alternatives, the single postulated event with the largest calculated radiologic impact for any waste class is an explosion of a mixture of ferri/ferro cyanide precipitates during the mechanical retrieval or microwave drying of the salt cake in single shell waste tanks. The anticipated downwind dose (70-year dose commitment) to the maximally exposed individual is 3 rem with a total population dose of 7000 man-rem. The same individual would receive 7 rem from natural background radiation during the same time period, and the same population would receive 3,000,000 man-rem. Radiological impacts to the public from all other postulated accidents would be less than that from this accident; furthermore, the radiological impacts resulting from this accident would be less than one-half that from the natural background radiation dose.

Mishima, J.; Sutter, S.L.; Hawley, K.A.; Jenkins, C.E.; Napier, B.A.

1986-02-01T23:59:59.000Z

60

Process for preparing liquid wastes  

DOE Patents [OSTI]

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

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

1997-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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

Waste minimization at a plutonium processing facility  

SciTech Connect (OSTI)

As part of Los Alamos National Laboratory`s (LANL) mission to reduce the nuclear danger throughout the world, the plutonium processing facility at LANL maintains expertise and skills in nuclear weapons technologies as well as leadership in all peaceful applications of plutonium technologies, including fuel fabrication for terrestrial and space reactors and heat sources and thermoelectric generators for space missions. Another near-term challenge resulted from two safety assessments performed by the Defense Nuclear Facilities Safety Board and the U.S. Department of Energy during the past two years. These assessments have necessitated the processing and stabilization of plutonium contained in tons of residues so that they can be stored safely for an indefinite period. This report describes waste streams and approaches to waste reduction of plutonium management.

Pillay, K.K.S. [Los Alamos National Laboratory, NM (United States)

1995-12-31T23:59:59.000Z

62

Department of Energy plan for recovery and utilization of nuclear byproducts from defense wastes. Volume 2  

SciTech Connect (OSTI)

Nuclear wastes from the defense production cycle contain many uniquely useful, intrinsically valuable, and strategically important materials. These materials have a wide range of known and potential applications in food technology, agriculture, energy, public health, medicine, industrial technology, and national security. Furthermore, their removal from the nuclear waste stream can facilitate waste management and yield economic, safety, and environmental advantages in the management and disposal of the residual nuclear wastes that have no redemptive value. This document is the program plan for implementing the recovery and beneficial use of these valuable materials. An Executive Summary of this document, DOE/DP-0013, Vol. 1, January 1983, is available. Program policy, goals and strategy are stated in Section 2. Implementation tasks, schedule and funding are detailed in Section 3. The remaining five sections and the appendixes provide necessary background information to support these two sections. Section 4 reviews some of the unique properties of the individual byproduct materials and describes both demonstrated and potential applications. The amounts of byproduct materials that are available now for research and demonstration purposes, and the amounts that could be recovered in the future for expanded applications are detailed in Section 5. Section 6 describes the effects byproduct recovery and utilization have on the management and final disposal of nuclear wastes. The institutional issues that affect the recovery, processing and utilization of nuclear byproducts are discussed in Section 7. Finally, Section 8 presents a generalized mathematical process by which applications can be evaluated and prioritized (rank-ordered) to provide planning data for program management.

Not Available

1983-08-01T23:59:59.000Z

63

Economic evaluation of volume reduction for Defense transuranic waste  

SciTech Connect (OSTI)

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

Brown, C.M.

1981-07-01T23:59:59.000Z

64

Silicate Based Glass Formulations for Immobilization of U.S. Defense Wastes Using Cold Crucible Induction Melters  

SciTech Connect (OSTI)

The cold crucible induction melter (CCIM) is an alternative technology to the currently deployed liquid-fed, ceramic-lined, Joule-heated melter for immobilizing of U.S. tank waste generated from defense related reprocessing. In order to accurately evaluate the potential benefits of deploying a CCIM, glasses must be developed specifically for that melting technology. Related glass formulation efforts have been conducted since the 1990s including a recent study that is first documented in this report. The purpose of this report is to summarize the silicate base glass formulation efforts for CCIM testing of U.S. tank wastes. Summaries of phosphate based glass formulation and phosphate and silicate based CCIM demonstration tests are reported separately (Day and Ray 2013 and Marra 2013, respectively). Combined these three reports summarize the current state of knowledge related to waste form development and process testing of CCIM technology for U.S. tank wastes.

Smith, Gary L.; Kim, Dong-Sang; Schweiger, Michael J.; Marra, James C.; Lang, Jesse B.; Crum, Jarrod V.; Crawford, Charles L.; Vienna, John D.

2014-05-22T23:59:59.000Z

65

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER  

SciTech Connect (OSTI)

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

G. Radulesscu; J.S. Tang

2000-06-07T23:59:59.000Z

66

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

67

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

68

Independent Oversight Assessment, Salt Waste Processing Facility...  

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

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

69

CHARACTERIZATION OF DEFENSE NUCLEAR WASTE USING HAZARDOUS WASTE GUIDANCE. APPLICATIONS TO HANFORD SITE ACCELERATED HIGH-LEVEL WASTE TREATMENT AND DISPOSAL MISSION0  

SciTech Connect (OSTI)

Federal hazardous waste regulations were developed for management of industrial waste. These same regulations are also applicable for much of the nation's defense nuclear wastes. At the U.S. Department of Energy's (DOE) Hanford Site in southeast Washington State, one of the nation's largest inventories of nuclear waste remains in storage in large underground tanks. The waste's regulatory designation and its composition and form constrain acceptable treatment and disposal options. Obtaining detailed knowledge of the tank waste composition presents a significant portion of the many challenges in meeting the regulatory-driven treatment and disposal requirements for this waste. Key in applying the hazardous waste regulations to defense nuclear wastes is defining the appropriate and achievable quality for waste feed characterization data and the supporting evidence demonstrating that applicable requirements have been met at the time of disposal. Application of a performance-based approach to demonstrating achievable quality standards will be discussed in the context of the accelerated high-level waste treatment and disposal mission at the Hanford Site.

Hamel, William; Huffman, Lori; Lerchen, Megan; Wiemers, Karyn

2003-02-27T23:59:59.000Z

70

Salt Waste Processing Initiatives  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September2-SCORECARD-01-24-13 Page 1 of 1 DepartmentSalt Waste1

71

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

SciTech Connect (OSTI)

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

72

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

2000-01-01T23:59:59.000Z

73

EIS-0113: Disposal of Hanford Defense High-Level, Transuranic and Tank Waste, Hanford Site, Richland, Washington  

Broader source: Energy.gov [DOE]

The U.S. Department of Energy developed this EIS to examine the potential environmental impacts of final disposal options for legacy and future radioactive defense wastes stored at the Hanford Site.

74

Waste minimization in semiconductor processing  

SciTech Connect (OSTI)

The US semiconductor industry uses 5--7 thousand pounds of arsine annually. Fifty to eighty percent of the arsine used becomes a waste product, which requires abatement. Traditional methods of abatement are reviewed with an emphasis on dry chemical scrubbing. A variety of dry chemical scrubbing materials were evaluated for arsine capacity, using activated carbon as the baseline for comparison. Of the available technologies, dry chemical scrubbing is the most effective means of minimizing arsenic containing waste generated from semiconductor effluents. A copper oxide based media has been identified which has high capacity, high efficiency and treats the spectrum of gases used in MOCVD processes. Reclaim and recovery of spent scrubber media has the potential to drastically reduce arsenic waste from semiconductor manufacturing.

Hardwick, S.J.; Mailloux, J.C. [Novapure Corp., Danbury, CT (United States)

1994-12-31T23:59:59.000Z

75

Tank Waste and Waste Processing | Department of Energy  

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

breakthrough immobilization technologies. Currently projects are focusing on: In-tank sludge washing at Hanford Enhanced waste processing at Idaho, Hanford, and Savannah River...

76

Thermal processing systems for TRU mixed waste  

SciTech Connect (OSTI)

This paper presents preliminary ex situ thermal processing system concepts and related processing considerations for remediation of transuranic (TRU)-contaminated wastes (TRUW) buried at the Radioactive Waste Management Complex (RWMC) of the Idaho National Engineering Laboratory (INEL). Anticipated waste stream components and problems are considered. Thermal processing conditions required to obtain a high-integrity, low-leachability glass/ceramic final waste form are considered. Five practical thermal process system designs are compared. Thermal processing of mixed waste and soils with essentially no presorting and using incineration followed by high temperature melting is recommended. Applied research and development necessary for demonstration is also recommended.

Eddy, T.L.; Raivo, B.D.; Anderson, G.L.

1992-01-01T23:59:59.000Z

77

Thermal processing systems for TRU mixed waste  

SciTech Connect (OSTI)

This paper presents preliminary ex situ thermal processing system concepts and related processing considerations for remediation of transuranic (TRU)-contaminated wastes (TRUW) buried at the Radioactive Waste Management Complex (RWMC) of the Idaho National Engineering Laboratory (INEL). Anticipated waste stream components and problems are considered. Thermal processing conditions required to obtain a high-integrity, low-leachability glass/ceramic final waste form are considered. Five practical thermal process system designs are compared. Thermal processing of mixed waste and soils with essentially no presorting and using incineration followed by high temperature melting is recommended. Applied research and development necessary for demonstration is also recommended.

Eddy, T.L.; Raivo, B.D.; Anderson, G.L.

1992-08-01T23:59:59.000Z

78

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

SciTech Connect (OSTI)

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

79

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 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

80

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

Note: This page contains sample records for the topic "defense waste processing" 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

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

SciTech Connect (OSTI)

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

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

1996-03-01T23:59:59.000Z

82

THE USE OF POLYMERS IN RADIOACTIVE WASTE PROCESSING SYSTEMS  

SciTech Connect (OSTI)

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

83

Evaluation of defense-waste glass produced by full-scale vitrification equipment  

SciTech Connect (OSTI)

Three full-scale vitrification processes at the Pacific Northwest Laboratory produced over 67,000 kg of simulated nuclear-waste glass from March 1979 to August 1980. Samples were analyzed to monitor process operation and evaluate the resulting glass product. These processes are: Spray Calciner/In-Can Melter (SC/ICM); Spray Calciner/Calcine-Fed Ceramic Melter (SC/CFCM); and Liquid-Fed Ceramic Melter (LFCM). Waste components in the process feed varied less than +- 10%. The SC/ICM and SC/CFCM which use separate waste and frit feed systems showed larger glass compositional variation than the LFCM, which processed only premixed feed during this period. The SC/ICM and SC/CFCM product contained significant amounts of acmite crystals, while the LFCM product was largely amorphous. In addition, the lower portion of all SC/ICM-filled canisters contained a zone rich in waste components. A product chemical durability as determined by pH4 and soxhlet leach tests varied considerably. Aside from increased durability under pH4 conditions with decreasing waste content, glass composition, microstructure and melting process did not correlate with glass durability. For all samples analyzed, the weight loss under pH4 conditions ranged from 17.7 to 85.2 wt %. Soxhlet conditions produced weight losses from 1.78 to 3.56 wt %.

Lukacs, J.M.; Petkus, L.L.; Mellinger, G.B.

1981-09-01T23:59:59.000Z

84

Alternate approaches to verifying the structural adequacy of the Defense High Level Waste Shipping Cask  

SciTech Connect (OSTI)

In the early 1980s, the US Department of Energy/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as one that fully complies with all applicable DOE, Nuclear Regulatory Commission (NRC), and Department of Transportation (DOT) regulations. General Atomics (GA) designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This topical report presents the results of the two analytical approaches and the model testing results. The purpose of this work is to show that there are two viable analytical alternatives to verify the structural adequacy of a Type B package and to obtain an NRC license. It addition, this data will help to support the future acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing.

Zimmer, A.; Koploy, M.

1991-12-01T23:59:59.000Z

85

Electrochemical/Pyrometallurgical Waste Stream Processing and Waste Form Fabrication  

SciTech Connect (OSTI)

This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration between US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and waste form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.

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

2014-12-01T23:59:59.000Z

86

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014 EIS-0474:November 16,8, 1983:

87

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

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

Transuranic Waste Processing Center, September 2013 Independent Oversight Review, Oak Ridge Transuranic Waste Processing Center, September 2013 September 2013 Review of Management...

88

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

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

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

89

Examples of Process Modifications that Reduce Waste  

E-Print Network [OSTI]

The ideal approach to reducing or eliminating waste products is to avoid making them in the first place. This article will examine numerous process modifications that have accomplished that goal. We'll look at changes to raw materials, reactors...

Nelson, K. E.

90

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

91

Exploratory study of complexant concentrate waste processing  

SciTech Connect (OSTI)

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

92

Waste Heat Management Options for Improving Industrial Process...  

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

of waste heat streams, and options for recovery including Combined Heat and Power. Waste Heat Management Options for Improving Industrial Process Heating Systems...

93

Department of Energy plan for recovery and utilization of nuclear byproducts from defense wastes. Volume 1. Executive summary  

SciTech Connect (OSTI)

Nuclear byproducts are a major national resource that has yet to be incorporated into the economy. The current Defense Byproducts Program is designed to match specific military and commercial needs with the availability of valuable products which are currently treated as waste at considerable expense in waste management costs. This program plan focuses on a few specific areas with the greatest potential for near-term development and application. It also recognizes the need for a continuing effort to develop new applications for byproducts and to continue to assess the impacts on waste management. The entire program has been, and will continue to be structured so as to ensure the safety of the public and maintain the purity of the environment. Social and institutional concerns have been recognized and will be handled appropriately. A significant effort will be undertaken to inform the public of the benefits of byproduct use and of the care being taken to ensure safe, efficient operation.

None

1983-08-01T23:59:59.000Z

94

Waste Minimization Study on Pyrochemical Reprocessing Processes  

SciTech Connect (OSTI)

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

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

2006-07-01T23:59:59.000Z

95

Waste minimization by process modification  

SciTech Connect (OSTI)

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

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

1993-01-01T23:59:59.000Z

96

Apply process integration to waste minimization  

SciTech Connect (OSTI)

This article presents a systematic method for identifying process modifications to minimize waste generation. It is based on the hierarchical decision procedure, which provides a framework for identifying process improvement options and evaluating heat and mass integration opportunities. The article deals specifically with an adaptation of the hierarchical decision approach for use in pollution abatement applications. The article also illustrates the use of the technique by applying it to the fluid catalytic cracking unit at Amoco Oil Co.'s Yorktown, VA, refinery.

Rossiter, A.P.; Spriggs, H.D. (Linnhoff March, Inc., Leesburg, VA (United States)); Klee, H. Jr. (Amoco Corp., Chicago, IL (United States))

1993-01-01T23:59:59.000Z

97

Process for treating alkaline wastes for vitrification  

DOE Patents [OSTI]

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

Hsu, C.L.W.

1995-07-25T23:59:59.000Z

98

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

99

Waste immobilization process development at the Savannah River Plant  

SciTech Connect (OSTI)

Processes to immobilize various wasteforms, including waste salt solution, transuranic waste, and low-level incinerator ash, are being developed. Wasteform characteristics, process and equipment details, and results from field/pilot tests and mathematical modeling studies are discussed.

Charlesworth, D L

1986-01-01T23:59:59.000Z

100

Integrated DWPF Melter System (IDMS) campaign report: Hanford Waste Vitrification Plan (HWVP) process demonstration  

SciTech Connect (OSTI)

Vitrification facilities are being developed worldwide to convert high-level nuclear waste to a durable glass form for permanent disposal. Facilities in the United States include the Department of Energy`s Defense Waste Processing Facility (DWPF) at the Savannah River Site, the Hanford Waste Vitrification Plant (HWVP) at the Hanford Site and the West Valley Demonstration Project (WVDP) at West Valley, NY. At each of these sites, highly radioactive defense waste will be vitrified to a stable borosilicate glass. The DWPF and WVDP are near physical completion while the HWVP is in the design phase. The Integrated DWPF Melter System (IDMS) is a vitrification test facility at the Savannah River Technology Center (SRTC). It was designed and constructed to provide an engineering-scale representation of the DWPF melter and its associated feed preparation and off-gas treatment systems. Because of the similarities of the DWPF and HWVP processes, the IDMS facility has also been used to characterize the processing behavior of a reference NCAW simulant. The demonstration was undertaken specifically to determine material balances, to characterize the evolution of offgas products (especially hydrogen), to determine the effects of noble metals, and to obtain general HWVP design data. The campaign was conducted from November, 1991 to February, 1992.

Hutson, N.D.

1992-08-10T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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

Keeping Track of the National Transuranic Program Complex Defense Transuranic Waste  

SciTech Connect (OSTI)

The long-term performance of the Waste Isolation Pilot Plant (WIPP) disposal system in southeastern New Mexico is assessed periodically using transuranic (TRU) waste physical and radiological properties and other information describing the waste. This TRU waste estimate is based on the best knowledge of the TRU waste across the DOE complex at the time repository performance is assessed. TRU waste inventory was collected from each of the Department of Energy (DOE) sites that generated TRU waste for the Compliance Certification Application (CCA) and subsequently for the Compliance Re-certification Application (CRA) in order to support the assessments that ultimately led to certification and re-certification of the WIPP. In each case, information was collected, stored and maintained in the Transuranic Waste Baseline Inventory Database (TWBID) that was used to generate tables describing the volumetric, physical, and radiological properties of the TRU waste. The tables and other descriptions of the waste were reported in baseline reports for the certification and the re-certification. Information maintained in the TWBID database has now been transferred to a new qualified database that utilizes a more efficient operating configuration. This database known as the Comprehensive Inventory Database (CID) will be the information repository for TRU waste destined to WIPP, and the source for information submitted for annual Transuranic Waste Inventory Update Reports to be used in future repository performance assessments (PAs) and re-certifications. The information that has been collected will support a wider range of data needs including waste management, transportation and strategic planning. (authors)

Crawford, B.; Lott, S.; McInroy, W.; VanSoest, G. [Los Alamos National Laboratory-Carlsbad Operations, Carlsbad, NM (United States); Patterson, R. [U.S. Department of Energy Carlsbad Field Office, Carlsbad, NM (United States)

2007-07-01T23:59:59.000Z

102

An approach for sampling solid heterogeneous waste at the Hanford Site waste receiving and processing and solid waste projects  

SciTech Connect (OSTI)

This paper addresses the problem of obtaining meaningful data from samples of solid heterogeneous waste while maintaining sample rates as low as practical. The Waste Receiving and Processing Facility, Module 1, at the Hanford Site in south-central Washington State will process mostly heterogeneous solid wastes. The presence of hazardous materials is documented for some packages and unknown for others. Waste characterization is needed to segregate the waste, meet waste acceptance and shipping requirements, and meet facility permitting requirements. Sampling and analysis are expensive, and no amount of sampling will produce absolute certainty of waste contents. A sampling strategy is proposed that provides acceptable confidence with achievable sampling rates.

Sexton, R.A.

1993-03-01T23:59:59.000Z

103

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

104

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

SciTech Connect (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

105

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

SciTech Connect (OSTI)

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

Wasan, Darsh T.

2007-10-09T23:59:59.000Z

106

Double Shell Tank (DST) Process Waste Sampling Subsystem Specification  

SciTech Connect (OSTI)

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

107

Buried waste integrated demonstration technology integration process  

SciTech Connect (OSTI)

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

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

1992-04-01T23:59:59.000Z

108

Buried waste integrated demonstration technology integration process  

SciTech Connect (OSTI)

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

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

1992-04-01T23:59:59.000Z

109

GEOTECHNICAL/GEOCHEMICAL CHARACTERIZATION OF ADVANCED COAL PROCESS WASTE STREAMS  

SciTech Connect (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

110

EIS-0023: Long-Term Management of Defense High-Level Radioactive Wastes (Research and Development Program for Immobilization), Savannah River Plant, Aiken, South Carolina  

Broader source: Energy.gov [DOE]

This environmental impact statement (EIS) analyzes the environmental implications of the proposed continuation of a large Federal research and development (R&D) program directed toward the immobilization of the high-level radioactive wastes resulting from chemical separations operations for defense radionuclides production at the DOE Savannah River Plant (SRP) near Aiken, South Carolina.

111

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

SciTech Connect (OSTI)

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

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

1993-12-31T23:59:59.000Z

112

Report on Separate Disposal of Defense High-Level Radioactive Waste |  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September 15,2015Department of Energy on Separate Disposal of Defense

113

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

E-Print Network [OSTI]

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

Reyle, Uwe

114

INVESTIGATION OF WASTE GLASS POURING PROCESS OVER A KNIFE EDGE  

SciTech Connect (OSTI)

Vitrification is the process of capturing radioactive waste in glass. The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF) is one of the facilities using the vitrification technology to treat and immobilize radioactive waste. The objective of the project is to investigate the pouring behavior of molten glass over a pour spout knife edge. Experiments are run using simulant glass containing the same chemical formulation as the radioactive sludge glass, but without radioactive contaminants. The purpose of these tests is to obtain actual glass data that, when combined with previous cold data from other fluids, will provide an overall understanding of the physics of liquids flowing over a pour spout and knife edge, A specific objective is to verify computational fluid dynamics (CFD) models with a range of liquid data with particular emphasis on glass so as to provide confidence in use of these CFD models for designing a new improved pour spout for the DWPF melter. The work to be performed at FIU-HCET includes assembling the melting and pouring system that mimics the DWPF melter and determining the key parameters that may influence wicking. Information from the FIU-HCET melter tests will lead to better operating guidelines for the DWPF melter so as to avoid wicking. During FY98, a bench-scale melter complete with pour spout and a knife edge was designed and assembled at FIU-HCET. Initially, the system was tested with glycerine. Subsequently, glass provided by SRS was used for experimentation. Flow visualization tests were performed with the melter in FY98 to investigate the pouring behavior of molten glass over a pour spout model simulating a DWPF pour spout of the original design. Simulant glass containing the same chemical formulation as sludge glass but without radioactive contaminants was used in the tests. All the tasks and milestones mentioned in the PTP for the project were accomplished. The project completed its second year, and this document reports the tasks and milestones that were accomplished during the 1998 fiscal year.

M.A. Ebadian, Ph.D.

1999-01-01T23:59:59.000Z

115

Characterization of industrial process waste heat and input heat streams  

SciTech Connect (OSTI)

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

116

Transuranic Waste Processing Center Oak Ridge Site Specific...  

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

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

117

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

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

Facility - December 2013 Independent Oversight Review, Oak Ridge Transuranic Waste Processing Facility - December 2013 December 2013 Review of the Fire Protection Program and Fire...

118

Waste receiving and processing facility module 1 auditable safetyanalysis  

SciTech Connect (OSTI)

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

Bottenus, R.J.

1997-02-01T23:59:59.000Z

119

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

120

Newly Generated Liquid Waste Processing Alternatives Study, Volume 1  

SciTech Connect (OSTI)

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

Note: This page contains sample records for the topic "defense waste processing" 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

Hydrothermal processing of Hanford tank wastes: Process modeling and control  

SciTech Connect (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

122

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

SciTech Connect (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

123

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

SciTech Connect (OSTI)

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

TRINER, G.C.

1999-11-01T23:59:59.000Z

124

Thermal processing system concepts and considerations for RWMC buried waste  

SciTech Connect (OSTI)

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

125

Materials selection for process equipment in the Hanford waste vitrification plant  

SciTech Connect (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

126

Sodium Bearing Waste Processing Alternatives Analysis  

SciTech Connect (OSTI)

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

127

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

SciTech Connect (OSTI)

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

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

1996-08-01T23:59:59.000Z

128

Chapter 38 Hazardous Waste Permitting Process (Kentucky)  

Broader source: Energy.gov [DOE]

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

129

Technical evaluation of proposed Ukrainian Central Radioactive Waste Processing Facility  

SciTech Connect (OSTI)

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

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

1996-06-01T23:59:59.000Z

130

Process development for remote-handled mixed-waste treatment  

SciTech Connect (OSTI)

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

131

Proposed Changes to EPA's Transuranic Waste Characterization Approval Process  

SciTech Connect (OSTI)

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

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

2003-02-25T23:59:59.000Z

132

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

133

Detection of localized and general corrosion of mild steel in simulated defense nuclear waste solutions using electrochemical noise analysis  

SciTech Connect (OSTI)

Underground waste tanks fabricated from mild steel store more than 60 million gallons of radioactive waste from 50 years of weapons production. Leaks are suspected in a significant number of tanks. The probable modes of corrosion failures are reported to be localized corrosion (e.g. nitrate stress corrosion cracking and pitting). The use of electrochemical noise (EN) for the monitoring and detection of localized corrosion processes has received considerable attention and application over the last several years. Proof of principle laboratory tests were conducted to verify the capability of EN evaluation to detect localized corrosion and to compare the predictions of general corrosion obtained from EN with those derived from other sources. Simple, pre-fabricated flat and U-bend specimens of steel alloys A516-Grade 60 (UNS K02100) and A537-CL 1 (UNS K02400) were immersed in temperature controlled simulated waste solutions. The simulated waste solution was either 5M NaNO{sub 3} with 0.3M NaOH at 90 C or 11M NaNO{sub 3} with 0.15M NaOH at 95 C. The electrochemical noise activity from the specimens was monitored and recorded for periods ranging between 140 and 240 hours. At the end of each test period, the specimens were metallographically examined to correlated EN data with corrosion damage.

Edgemon, G.L.; Ohl, P.C. [Westinghouse Hanford Co., Richland, WA (United States); Bell, G.E.C. [M.J. Schiff and Associates, Inc., Claremont, CA (United States); Wilson, D.F. [Oak Ridge National Lab., TN (United States)

1995-12-01T23:59:59.000Z

134

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

SciTech Connect (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

135

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).

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

1989-01-01T23:59:59.000Z

136

Tank Waste and Waste Processing | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security AdministrationcontrollerNanocrystallineForeign ObjectOUR8, 2013 FINAL MEETING SUMMARY7,Tank Waste and

137

Process Knowledge Summary Report for Advanced Test Reactor Complex Contact-Handled Transuranic Waste Drum TRA010029  

SciTech Connect (OSTI)

This Process Knowledge Summary Report summarizes information collected to satisfy the transportation and waste acceptance requirements for the transfer of one drum containing contact-handled transuranic (TRU) actinide standards generated by the Idaho National Laboratory at the Advanced Test Reactor (ATR) Complex to the Advanced Mixed Waste Treatment Project (AMWTP) for storage and subsequent shipment to the Waste Isolation Pilot Plant for final disposal. The drum (i.e., Integrated Waste Tracking System Bar Code Number TRA010029) is currently stored at the Materials and Fuels Complex. The information collected includes documentation that addresses the requirements for AMWTP and applicable sections of their Resource Conservation and Recovery Act permits for receipt and disposal of this TRU waste generated from ATR. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for this TRU waste originating from ATR.

B. R. Adams; R. P. Grant; P. R. Smith; J. L. Weisgerber

2013-09-01T23:59:59.000Z

138

Description of processes for the immobilization of selected transuranic wastes  

SciTech Connect (OSTI)

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

Timmerman, C.L.

1980-12-01T23:59:59.000Z

139

Hanford's Simulated Low Activity Waste Cast Stone Processing  

SciTech Connect (OSTI)

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, Young

2013-08-20T23:59:59.000Z

140

Co-processing of agricultural and biomass waste with coal  

SciTech Connect (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

Note: This page contains sample records for the topic "defense waste processing" 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

E-Print Network 3.0 - agriculture process waste Sample Search...  

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

Agricultural and food processing plants can turn waste into power can turn... . Wastewater treatment plants (WWTPs), waste streams from food and beverage processing plants,...

142

Building 251 Radioactive Waste Characterization by Process Knowledge  

SciTech Connect (OSTI)

Building 251 is the Lawrence Livermore National Laboratory Heavy Elements Facility. Operations that involved heavy elements with uncontained radioisotopes including transuranic elements took place inside of glove boxes and fume hoods. These operations included process and solution chemistry, dissolutions, titrations, centrifuging, etc., and isotope separation. Operations with radioactive material which presently take place outside of glove boxes include storage, assaying, packing and unpacking and inventory verification. Wastes generated inside glove boxes will generally be considered TRU or Greater Than Class C (GTCC). Wastes generated in the RMA, outside glove boxes, is presumed to be low level waste. This process knowledge quantification method may be applied to waste generated anywhere within or around B251. The method is suitable only for quantification of waste which measures below the MDA of the Blue Alpha meter (i.e. only material which measures as Non-Detect with the blue alpha is to be characterized by this method).

Dominick, J L

2002-05-29T23:59:59.000Z

143

Technical resource document for assured thermal processing of wastes  

SciTech Connect (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

144

Centralized processing of contact-handled TRU waste feasibility analysis  

SciTech Connect (OSTI)

This report presents work for the feasibility study of central processing of contact-handled TRU waste. Discussion of scenarios, transportation options, summary of cost estimates, and institutional issues are a few of the subjects discussed. (JDL)

Not Available

1986-12-01T23:59:59.000Z

145

The Insanity of the Mens Rea Model: Due Process and the Abolition of the Insanity Defense  

E-Print Network [OSTI]

In the last 15 years a flurry of legislative activity has taken place as states have attempted to redefine the insanity defense. This article focuses on those states who chose not just to refine the definition of insanity, ...

Gilles Phillips, Jean K.; Woodman, Rebecca E.

2007-01-01T23:59:59.000Z

146

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

147

Defense-in-Depth, How Department of Energy Implements Radiation...  

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

Defense-in-Depth, How Department of Energy Implements Radiation Protection in Low Level Waste Disposal Defense-in-Depth, How Department of Energy Implements Radiation Protection in...

148

Accelerator Production of Tritium project process waste assessment  

SciTech Connect (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

149

Wastes from plutonium conversion and scrap recovery operations  

SciTech Connect (OSTI)

This report deals with the handling of defense-related wastes associated with plutonium processing. It first defines the different waste categories along with the techniques used to assess waste content. It then discusses the various treatment approaches used in recovering plutonium from scrap. Next, it addresses the various waste management approaches necessary to handle all wastes. Finally, there is a discussion of some future areas for processing with emphasis on waste reduction. 91 refs., 25 figs., 4 tabs.

Christensen, D.C.; Bowersox, D.F.; McKerley, B.J.; Nance, R.L.

1988-03-01T23:59:59.000Z

150

Electromagnetic mixed waste processing system for asbestos decontamination  

SciTech Connect (OSTI)

DOE sites contain a broad spectrum of asbestos materials (cloth, pipe lagging, sprayed insulation and other substances) which are contaminated with a combination of hazardous and radioactive wastes due to its use during the development of the US nuclear weapons complex. These wastes consist of cutting oils, lubricants, solvents, PCBs, heavy metals and radioactive contaminants. The radioactive contaminants are the activation, decay, and fission products of DOE operations. To allow disposal, the asbestos must be converted chemically, followed by removing and separating the hazardous and radioactive materials to prevent the formation of mixed wastes and to allow for both sanitary disposal and effective decontamination. Currently, no technology exists that can meet these sanitary and other objectives. An attempt was made to apply techniques that have already proved successful in the mining, oil, and metals processing industries to the development of a multi-stage process to remove and separate hazardous chemical radioactive materials from asbestos. This process uses three methods: ABCOV chemicals which converts the asbestos to a sanitary waste; dielectric heating to volatilize the organic materials; and electrochemical processing for the removal of heavy metals, RCRA wastes and radionuclides. This process will result in the destruction of over 99% of the asbestos; limit radioactive metal contamination to 0.2 Bq alpha per gram and 1 Bq beta and gamma per gram; reduce hazardous organics to levels compatible with current EPA policy for RCRA delisting; and achieve TCLP limits for all solidified waste.

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

1994-12-31T23:59:59.000Z

151

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

SciTech Connect (OSTI)

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

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

2013-08-29T23:59:59.000Z

152

Process Simulation as Applied to Transuranic Waste Management  

SciTech Connect (OSTI)

The National Transuranic Waste System Model (the Model) is a computer simulation designed to evaluate the preparation and flow of TRU waste from generator sites throughout the Department of Energy (the Department) complex to the Waste Isolation Pilot Plant (WIPP) facility for disposal. The Model uses process simulation software to predict waste outputs of waste management operations as a function of time over the life of the WIPP. Process simulation modeling is a tool used by many industries, both private and public, to evaluate complex systems. For example a manufacturing plant might use process simulation to determine the possible effects of increasing the rate of production: will there be adequate resources (labor pool, raw goods, transportation capability); can the new production rate be sustained for an indefinite period of time without adding additional infrastructure. Process simulation modeling is also used by various military branches to ensure adequate supplies are delivered in a timely manner. The Department currently uses this technique as the basis for its National TRU Waste Management Plan Rev. 1 (DOE, 1997).

Brown, M.; Downes, S.; Trone, J.

1999-01-01T23:59:59.000Z

153

Defense Gallery  

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

Defense Gallery Inside the Museum Exhibitions Norris Bradbury Museum Lobby Defense Gallery Research Gallery History Gallery TechLab Virtual Exhibits invisible utility element...

154

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

155

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

156

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

SciTech Connect (OSTI)

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

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

1993-04-01T23:59:59.000Z

157

Waste Receiving and Processing Facility - Hanford Site  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLasDelivered energy consumption byAbout Printable VersionProtective ActionsWaste

158

TECHNOLOGY SUMMARY ADVANCING TANK WASTE RETRIEVAL AND PROCESSING  

SciTech Connect (OSTI)

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

159

TECHNOLOGY SUMMARY ADVANCING TANK WASTE RETREIVAL AND PROCESSING  

SciTech Connect (OSTI)

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

160

Crystallization process to reduce NORM-containing waste  

SciTech Connect (OSTI)

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

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

1995-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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

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

SciTech Connect (OSTI)

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

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

1994-12-01T23:59:59.000Z

162

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

SciTech Connect (OSTI)

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

163

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

SciTech Connect (OSTI)

A polyethylene encapsulation system for treatment of low-level radioactive, hazardous, and mixed wastes has been developed at Brookhaven National Laboratory. Polyethylene has several advantages compared with conventional solidification/stabilization materials such as hydraulic cements. Waste can be encapsulated with greater efficiency and with better waste form performance than is possible with hydraulic cement. The properties of polyethylene relevant to its long-term durability in storage and disposal environments are reviewed. Response to specific potential failure mechanisms including biodegradation, radiation, chemical attack, flammability, environmental stress cracking, and photodegradation are examined. These data are supported by results from extensive waste form performance testing including compressive yield strength, water immersion, thermal cycling, leachability of radioactive and hazardous species, irradiation, biodegradation, and flammability. The bench-scale process has been successfully tested for application with a number of specific problem'' waste streams. Quality assurance and performance testing of the resulting waste form confirmed scale-up feasibility. Use of this system at Rocky Flats Plant can result in over 70% fewer drums processed and shipped for disposal, compared with optimal cement formulations. Based on the current Rocky Flats production of nitrate salt per year, polyethylene encapsulation can yield an estimated annual savings between $1.5 million and $2.7 million, compared with conventional hydraulic cement systems. 72 refs., 23 figs., 16 tabs.

Kalb, P.D.; Heiser, J.H. III; Colombo, P.

1991-07-01T23:59:59.000Z

164

EIS-0062: Double-Shell Tanks for Defense High Level Waste Storage, Savannah River Site, Aiken, SC  

Broader source: Energy.gov [DOE]

This EIS analyzes the impacts of the various design alternatives for the construction of fourteen 1.3 million gallon high-activity radioactive waste tanks. The EIS further evaluates the effects of these alternative designs on tank durability, on the ease of waste retrieval from such tanks, and the choice of technology and timing for long-term storage or disposal of the wastes.

165

Field study of disposed solid wastes from advanced coal processes  

SciTech Connect (OSTI)

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

Not Available

1992-01-01T23:59:59.000Z

166

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

E-Print Network [OSTI]

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

Hasanbeigi, Ali

2013-01-01T23:59:59.000Z

167

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...

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

1980-01-01T23:59:59.000Z

168

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...

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

1980-01-01T23:59:59.000Z

169

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

170

Process for recovery of palladium from nuclear fuel reprocessing wastes  

DOE Patents [OSTI]

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

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

1980-06-16T23:59:59.000Z

171

Waste Processing Annual Technology Development Report 2007  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergy Cooperation | Department of EnergyDepartmentEnergyU.S. DOE EnvironmentalProcessing

172

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

SciTech Connect (OSTI)

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

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

1998-05-01T23:59:59.000Z

173

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

174

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

175

Process for the recovery of curium-244 from nuclear waste  

SciTech Connect (OSTI)

A process has been designed for the recovery of curium from purex waste. Curium and americium are separated from the lanthanides by a TALSPEAK extraction process using differential extraction. Equations were derived for the estimation of the economically optimum conditions for the extraction using laboratory batch extraction data. The preparation of feed for the extraction involves the removal of nitric acid from the Purex waste by vaporization under reduced pressure, the leaching of soluble nitrates from the resulting cake, and the oxalate precipitation of a pure lanthanide-actinide fraction. Final separation of the curium from americium is done by ion-exchange. The steps of the process, except ion-exchange, were tested on a laboratory scale and workable conditions were determined.

Posey, J.C.

1980-10-01T23:59:59.000Z

176

Process modeling of hydrogen production from municipal solid waste  

SciTech Connect (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

177

An Effective Waste Management Process for Segregation and Disposal of Legacy Mixed Waste at Sandia National Laboratories/New Mexico  

SciTech Connect (OSTI)

Sandia National Laboratories/New Mexico (SNL/NM) is a research and development facility that generates many highly diverse, low-volume mixed waste streams. Under the Federal Facility Compliance Act, SNL/NM must treat its mixed waste in storage to meet the Land Disposal Restrictions treatment standards. Since 1989, approximately 70 cubic meters (2500 cubic feet) of heterogeneous, poorly characterized and inventoried mixed waste was placed in storage that could not be treated as specified in the SNL/NM Site Treatment Plan. A process was created to sort the legacy waste into sixteen well- defined, properly characterized, and precisely inventoried mixed waste streams (Treatability Groups) and two low-level waste streams ready for treatment or disposal. From June 1995 through September 1996, the entire volume of this stored mixed waste was sorted and inventoried through this process. This process was planned to meet the technical requirements of the sorting operation and to identify and address the hazards this operation presented. The operations were routinely adapted to safely and efficiently handle a variety of waste matrices, hazards, and radiological conditions. This flexibility was accomplished through administrative and physical controls integrated into the sorting operations. Many Department of Energy facilities are currently facing the prospect of sorting, characterizing, and treating a large inventory of mixed waste. The process described in this paper is a proven method for preparing a diverse, heterogeneous mixed waste volume into segregated, characterized, inventoried, and documented waste streams ready for treatment or disposal.

Hallman, Anne K. [Sandia National Labs., Albuquerque, NM (United States); Meyer, Dann [IT Corporation, Albuquerque, NM (United States); Rellergert, Carla A. [Roy F. Weston, Inc., Albuquerque, NM (United States); Schriner, Joseph A. [Automated Solutions of Albuquerque, Albuquerque, NM (United States)

1998-06-01T23:59:59.000Z

178

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

179

The Hybrid Treatment Process for mixed radioactive and hazardous waste treatment  

SciTech Connect (OSTI)

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

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

1992-06-01T23:59:59.000Z

180

The Hybrid Treatment Process for mixed radioactive and hazardous waste treatment  

SciTech Connect (OSTI)

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

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

1992-06-01T23:59:59.000Z

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181

An effective waste management process for segregation and disposal of legacy mixed waste at Sandia National Laboratories/New Mexico  

SciTech Connect (OSTI)

Sandia National Laboratories/New Mexico (SNL/NM) is a research and development facility that generates many highly diverse, low-volume mixed waste streams. Under the Federal Facility Compliance Act, SNL/NM must treat its mixed waste in storage to meet the Land Disposal Restrictions treatment standards. Since 1989, approximately 70 cubic meters (2,500 cubic feet) of heterogeneous, poorly characterized and inventoried mixed waste was placed in storage that could not be treated as specified in the SNL/NM Site Treatment Plan. A process was created to sort the legacy waste into sixteen well-defined, properly characterized, and accurately inventoried mixed waste streams (Treatability Groups) and two low-level waste streams ready for treatment or disposal. From June 1995 through September 1996, the entire volume of this stored mixed waste was sorted and inventoried. This process was planned to meet the technical requirements of the sorting operation and to identify and address the hazards this operation presented. The operations were routinely adapted to safely and efficiently handle a variety of waste matrices, hazards, and radiological conditions. This flexibility was accomplished through administrative and physical controls integrated into the sorting operations. Many Department of Energy facilities are currently facing the prospect of sorting, characterizing, and treating a large inventory of mixed waste. The process described in this report is a proven method for preparing a diverse, heterogeneous mixed waste volume into segregated, characterized, inventoried, and documented waste streams ready for treatment or disposal.

Hallman, A.K. [Sandia National Labs., Albuquerque, NM (United States); Meyer, D. [IT Corp., Albuquerque, NM (United States); Rellergert, C.A. [Roy F. Weston, Inc., Albuquerque, NM (United States); Schriner, J.A. [Automated Solutions of Albuquerque, Inc., NM (United States)

1998-04-01T23:59:59.000Z

182

Record of Decision; Defense Waste Processing Facility at the Savannah River Site  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAXBalancedDepartment ofColumbusReport # INL/EXT-06-11478 ControlEIS | Department07889

183

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

SciTech Connect (OSTI)

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

184

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

185

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

SciTech Connect (OSTI)

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

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

1990-09-18T23:59:59.000Z

186

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

SciTech Connect (OSTI)

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

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

2003-01-01T23:59:59.000Z

187

Waste-heat recovery in batch processes using heat storage  

SciTech Connect (OSTI)

The waste-heat recovery in batch processes has been studied using the pinch-point method. The aim of the work has been to investigate theoretical and practical approaches to the design of heat-exchanger networks, including heat storage, for waste-heat recovery in batch processes. The study is limited to the incorporation of energy-storage systems based on fixed-temperature variable-mass stores. The background for preferring this to the alternatives (variable-temperature fixed-mass and constant-mass constant-temperature (latent-heat) stores) is given. It is shown that the maximum energy-saving targets as calculated by the pinch-point method (time average model, TAM) can be achieved by locating energy stores at either end of each process stream. This theoretically large number of heat-storage tanks (twice the number of process streams) can be reduced to just a few tanks. A simple procedure for determining a number of heat-storage tanks sufficient to achieve the maximum energy-saving targets as calculated by the pinch-point method is described. This procedure relies on combinatorial considerations, and could therefore be labeled the combinatorial method for incorporation of heat storage in heat-exchanger networks. Qualitative arguments justifying the procedure are presented. For simple systems, waste-heat recovery systems with only three heat-storage temperatures (a hot storage, a cold storage, and a heat store at the pinch temperature) often can achieve the maximum energy-saving targets. Through case studies, six of which are presented, it is found that a theoretically large number of heat-storage tanks (twice the number of process streams) can be reduced to just a few tanks. The description of these six cases is intended to be sufficiently detailed to serve as benchmark cases for development of alternative methods.

Stoltze, S.; Mikkelsen, J.; Lorentzen, B.; Petersen, P.M.; Qvale, B. [Technical Univ. of Denmark, Lyngby (Denmark). Lab. for Energetics

1995-06-01T23:59:59.000Z

188

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

SciTech Connect (OSTI)

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

189

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

SciTech Connect (OSTI)

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

None

2010-05-01T23:59:59.000Z

190

Process for treating waste water having low concentrations of metallic contaminants  

DOE Patents [OSTI]

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

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

2014-12-16T23:59:59.000Z

191

Method for co-processing waste rubber and carbonaceous material  

DOE Patents [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

192

Co-processing of agriculture and biomass waste with coal  

SciTech Connect (OSTI)

Biomass and bio-processed waste are potential candidates for co-liquefaction with coal. Specific materials used here include sawdust and poultry manure. Liquefaction experiments were run on each of these materials, separately and with coal, using tetralin as solvent at 350{degrees}C and 1000 psi(cold) hydrogen pressure for 1h. Total conversion was monitored, as well as conversion to asphaltenes, oils and gases. All the biomass samples are converted to oils and gases under the reaction conditions. Poultry manure seems to convert coal more completely, and to produce more oils and gases, than conventional liquefaction.

Stiller, A.H.; Dadyburjor, D.B.; Wann, J.P. [West Virginia Univ., Morgantown, WV (United States)

1995-12-01T23:59:59.000Z

193

Nuclear waste management. Semiannual progress report, October 1983-March 1984  

SciTech Connect (OSTI)

Progress in the following studies on radioactive waste management is reported: defense waste technology; Nuclear Waste Materials Characterization Center; waste isolation; and supporting studies. 58 figures, 22 tables.

McElroy, J.L.; Powell, J.A.

1984-06-01T23:59:59.000Z

194

A Cask Processing Enclosure for the TRU Waste Processing Center - 13408  

SciTech Connect (OSTI)

This paper will discuss the key elements considered in the design, construction, and use of an enclosure system built for the TRU Waste Processing Center (TWPC). The TWPC system is used for the repackaging and volume reduction of items contaminated with radioactive material, hazardous waste and mixed waste. The modular structural steel frame and stainless steel skin was designed for rapid field erection by the use of interchangeable self-framing panel sections to allow assembly of a sectioned containment building and for ease of field mobility. The structure was installed on a concrete floor inside of an outer containment building. The major sections included an Outer Cask Airlock, Inner Cask Airlock, Cask Process Area, and Personnel Airlocks. Casks in overpacks containing transuranic waste are brought in via an inter-site transporter. The overpack lid is removed and the cask/overpack is transferred into the Outer Cask Airlock. A contamination cover is installed on the overpack body and the Outer Cask Airlock is closed. The cask/overpack is transferred into the Inner Cask Airlock on a cask bogie and the Inner Cask Airlock is closed. The cask lid is removed and the cask is transferred into the Cask Process Area where it is placed on a cask tilting station. Once the Cask Processing Area is closed, the cask tilt station is activated and wastes are removed, size reduced, then sorted and re-packaged into drums and standard waste boxes through bag ports. The modular system was designed and built as a 'Fast Track' project at IP Systems in Broomfield Colorado and then installed and is currently in use at the DOE TWPC located near Oak Ridge, Tennessee. (authors)

Newman, John T.; Mendez, Nicholas [IP Systems, Inc., 2685 Industrial Lane, Broomfield, Colorado 80020 (United States)] [IP Systems, Inc., 2685 Industrial Lane, Broomfield, Colorado 80020 (United States)

2013-07-01T23:59:59.000Z

195

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

SciTech Connect (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

196

Activated sludge process: Waste treatment. (Latest citations from the Biobusiness database). Published Search  

SciTech Connect (OSTI)

The bibliography contains citations concerning the use of the activated sludge process in waste and wastewater treatment. Topics include biochemistry of the activated sludge process, effects of various pollutants on process activity, effects of environmental variables such as oxygen and water levels, and nutrient requirements of microorganisms employed in activated sludge processes. The citations also explore use of the process to treat specific wastes, such as halocarbons, metallic wastes, and petrochemical effluents; and wastes from pharmaceutical and dairy processes. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

NONE

1996-01-01T23:59:59.000Z

197

Tank waste remediation system process engineering instruction manual  

SciTech Connect (OSTI)

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

198

An evaluation of neutralization for processing sodium-bearing liquid waste  

SciTech Connect (OSTI)

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

199

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

SciTech Connect (OSTI)

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

200

Rock mechanics contributions from defense programs  

SciTech Connect (OSTI)

An attempt is made at illustrating the many contributions to rock mechanics from US defense programs, over the past 30-plus years. Large advances have been achieved in the technology-base area covering instrumentation, material properties, physical modeling, constitutive relations and numerical simulations. In the applications field, much progress has been made in understanding and being able to predict rock mass behavior related to underground explosions, cratering, projectile penetration, and defense nuclear waste storage. All these activities stand on their own merit as benefits to national security. But their impact is even broader, because they have found widespread applications in the non-defense sector; to name a few: the prediction of the response of underground structures to major earthquakes, the physics of the earth`s interior at great depths, instrumentation for monitoring mine blasting, thermo-mechanical instrumentation useful for civilian nuclear waste repositories, dynamic properties of earthquake faults, and transient large-strain numerical modeling of geological processes, such as diapirism. There is not pretense that this summary is exhaustive. It is meant to highlight success stories representative of DOE and DOD geotechnical activities, and to point to remaining challenges.

Heuze, F.E.

1992-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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

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

202

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

203

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

204

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

205

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

206

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

207

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

Broader source: Energy.gov [DOE]

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

208

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

SciTech Connect (OSTI)

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

Fellinger, A.

2009-10-15T23:59:59.000Z

209

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

Broader source: Energy.gov [DOE]

Evaluation to determine whether Salt Waste Processing Facility Construction Project is continuing to perform at a level deserving DOE-VPP Star recognition.

210

Waste receiving and processing plant control system; system design description  

SciTech Connect (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

211

Verification Of The Defense Waste Processing Facility's (DWPF) Process Digestion Methods For The Sludge Batch 8 Qualification Sample  

SciTech Connect (OSTI)

This report contains the results and comparison of data generated from inductively coupled plasma – atomic emission spectroscopy (ICP-AES) analysis of Aqua Regia (AR), Sodium Peroxide/Sodium Hydroxide Fusion Dissolution (PF) and Cold Chem (CC) method digestions and Cold Vapor Atomic Absorption analysis of Hg digestions from the DWPF Hg digestion method of Sludge Batch 8 (SB8) Sludge Receipt and Adjustment Tank (SRAT) Receipt and SB8 SRAT Product samples. The SB8 SRAT Receipt and SB8 SRAT Product samples were prepared in the SRNL Shielded Cells, and the SRAT Receipt material is representative of the sludge that constitutes the SB8 Batch or qualification composition. This is the sludge in Tank 51 that is to be transferred into Tank 40, which will contain the heel of Sludge Batch 7b (SB7b), to form the SB8 Blend composition.

Click, D. R.; Edwards, T. B.; Wiedenman, B. J.; Brown, L. W.

2013-03-18T23:59:59.000Z

212

Zone Freezing Study for Pyrochemical Process Waste Minimization  

SciTech Connect (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

213

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

SciTech Connect (OSTI)

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

DEROSA, D.C.

2000-01-13T23:59:59.000Z

214

Processing and waste disposal representative for fusion breeder blanket systems  

SciTech Connect (OSTI)

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

Finn, P.A.; Vogler, S.

1987-01-01T23:59:59.000Z

215

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

SciTech Connect (OSTI)

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

216

Advanced thermal processing alternatives for solid waste management  

SciTech Connect (OSTI)

The 1990`s have seen a resurgence of interest in the development of new thermal processing alternatives for municipal solid waste (MSW). Sparked by increasingly stringent environmental regulations, much of this creative energy has been applied to technologies for the gasification of MSW: converting the solid, hard to handle material into a clean, medium to high-Btu fuel gas. Other developers have focussed on full combustion technology but with a {open_quotes}twist{close_quotes} that lowers emissions or reduces cost. A comprehensive study of these new technologies was recently completed under the sponsorship of the National Renewable Energy Laboratory of the U.S. Department of Energy. The study characterized the state-of-the-art among emerging MSW thermal processing technologies that have reached the point of `incipient commercialization.` More than 45 technologies now under development were screened to develop a short list of seven processes that have passed through the idea stage, laboratory and benchscale testing, and have been prototyped at an MSW feed rate of at least several tons per hour. In-depth review of these seven included inspections of operating pilot or prototype units and a detailed analysis of technical, environmental and economic feasibility issues. No attempt was made to select `the best` technology since best can only be defined in the context of the constraints, aspirations and circumstances of a specific, local situation. The basic flowsheet, heat and material balances and available environmental data were summarized to help the reader grasp the underlying technical concepts and their embodiment in hardware. Remaining development needs, as seen by the study team are presented. Economic analysis shows the general balance of capital and operating costs.

Niessen, W.R. [Camp Dresser & McKee Inc., Cambridge, MA (United States)

1997-12-01T23:59:59.000Z

217

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

218

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

SciTech Connect (OSTI)

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

219

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

SciTech Connect (OSTI)

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

220

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

SciTech Connect (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

Note: This page contains sample records for the topic "defense waste processing" 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

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

SciTech Connect (OSTI)

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

222

Corrosion and failure processes in high-level waste tanks  

SciTech Connect (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

223

Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel  

SciTech Connect (OSTI)

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

224

Management of salt waste from electrochemical processing of used nuclear fuel  

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

225

Nuclear Waste Management. Semiannual progress report, October 1984-March 1985  

SciTech Connect (OSTI)

Progress reports are presented for the following studies on radioactive waste management: defense waste technology; nuclear waste materials characterization center; and supporting studies. 19 figs., 29 tabs.

McElroy, J.L.; Powell, J.A. (comps.)

1985-06-01T23:59:59.000Z

226

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

Energy Savers [EERE]

objective of the Liquid Waste contract is to achieve closure of the SRS liquid waste tanks in compliance with the Federal Facilities Agreement, utilizing the Defense Waste...

227

Program Management at the National Nuclear Security Administration Office of Defense Nuclear Security: A Review of Program Management Documents and Underlying Processes  

SciTech Connect (OSTI)

The scope of this paper is to review the National Nuclear Security Administration Office of Defense Nuclear Security (DNS) program management documents and to examine the underlying processes. The purpose is to identify recommendations for improvement and to influence the rewrite of the DNS Program Management Plan (PMP) and the documentation supporting it. As a part of this process, over 40 documents required by DNS or its stakeholders were reviewed. In addition, approximately 12 other documents produced outside of DNS and its stakeholders were reviewed in an effort to identify best practices. The complete list of documents reviewed is provided as an attachment to this paper.

Madden, Michael S.

2010-05-01T23:59:59.000Z

228

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, X.; Einziger, R.E.

1997-01-28T23:59:59.000Z

229

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

230

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, X.; Einziger, R.E.

1997-08-12T23:59:59.000Z

231

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

SciTech Connect (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

232

Identification of existing waste heat recovery and process improvement technologies  

SciTech Connect (OSTI)

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

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

1984-03-01T23:59:59.000Z

233

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

SciTech Connect (OSTI)

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

234

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

SciTech Connect (OSTI)

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

235

Process of Integrating Screening and Detailed Risk-based Modeling Analyses to Ensure Consistent and Scientifically Defensible Results  

SciTech Connect (OSTI)

To support cleanup and closure of these tanks, modeling is performed to understand and predict potential impacts to human health and the environment. Pacific Northwest National Laboratory developed a screening tool for the United States Department of Energy, Office of River Protection that estimates the long-term human health risk, from a strategic planning perspective, posed by potential tank releases to the environment. This tool is being conditioned to more detailed model analyses to ensure consistency between studies and to provide scientific defensibility. Once the conditioning is complete, the system will be used to screen alternative cleanup and closure strategies. The integration of screening and detailed models provides consistent analyses, efficiencies in resources, and positive feedback between the various modeling groups. This approach of conditioning a screening methodology to more detailed analyses provides decision-makers with timely and defensible information and increases confidence in the results on the part of clients, regulators, and stakeholders.

Buck, John W.; McDonald, John P.; Taira, Randal Y.

2002-11-01T23:59:59.000Z

236

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

SciTech Connect (OSTI)

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

237

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

SciTech Connect (OSTI)

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

238

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

SciTech Connect (OSTI)

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

GR Golcar; NG Colton; JG Darab; HD Smith

2000-04-04T23:59:59.000Z

239

Macroencapsulation of low-level debris waste with the phosphate ceramic process  

SciTech Connect (OSTI)

Across the DOE complex, large quantities of contaminated debris and irradiated lead bricks require disposal. The preferred method for disposing of these wastes is macroencapsulation under U.S. Environmental Protection Agency Alternative Treatment Standards. Chemically bonded phosphate ceramics serve as a novel binder, developed at Argonne National Laboratory, for stabilizing and solidifying various low-level mixed wastes. Extremely strong, dense, and impervious to water intrusion, this material was developed with support from the U.S. Department of Energy`s Office of Science and Technology (DOE OST). In this investigation, CBPCs have been used to demonstrate macroencapsulation of various contaminated debris wastes, including cryofractured debris, lead bricks, and lead-lined plastic gloves. This paper describes the processing steps for fabricating the waste forms and the results of various characterizations performed on the waste forms. The conclusion is that simple and low-cost CBPCs are excellent material systems for macroencapsulating debris wastes.

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

1997-03-01T23:59:59.000Z

240

A Robust Power Remote Manipulator for Use in Waste Sorting, Processing, and Packaging - 12158  

SciTech Connect (OSTI)

Disposition of radioactive waste is one of the Department of Energy's (DOE's) highest priorities. A critical component of the waste disposition strategy is shipment of Transuranic (TRU) waste from DOE's Oak Ridge Reservation to the Waste Isolation Plant Project (WIPP) in Carlsbad, New Mexico. This is the mission of the DOE TRU Waste Processing Center (TWPC). The remote-handled TRU waste at the Oak Ridge Reservation is currently in a mixed waste form that must be repackaged in to meet WIPP Waste Acceptance Criteria (WAC). Because this remote-handled legacy waste is very diverse, sorting, size reducing, and packaging will require equipment flexibility and strength that is not possible with standard master-slave manipulators. To perform the wide range of tasks necessary with such diverse, highly contaminated material, TWPC worked with S.A. Technology (SAT) to modify SAT's Power Remote Manipulator (PRM) technology to provide the processing center with an added degree of dexterity and high load handling capability inside its shielded cells. TWPC and SAT incorporated innovative technologies into the PRM design to better suit the operations required at TWPC, and to increase the overall capability of the PRM system. Improving on an already proven PRM system will ensure that TWPC gains the capabilities necessary to efficiently complete its TRU waste disposition mission. The collaborative effort between TWPC and S.A. Technology has yielded an extremely capable and robust solution to perform the wide range of tasks necessary to repackage TRU waste containers at TWPC. Incorporating innovative technologies into a proven manipulator system, these PRMs are expected to be an important addition to the capabilities available to shielded cell operators. The PRMs provide operators with the ability to reach anywhere in the cell, lift heavy objects, perform size reduction associated with the disposition of noncompliant waste. Factory acceptance testing of the TWPC Powered Remote Manipulators has completed at SAT's Colorado facility, and on-site training at TWPC is scheduled to start in early 2012. (authors)

Cole, Matt; Martin, Scott [S.A. Technology, Loveland, Colorado 80537, Transuranic Waste Processing Center, Lenoir City, TN 37771 (United States)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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

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

SciTech Connect (OSTI)

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

Benjamin, M.M.

1999-10-01T23:59:59.000Z

242

Formulation and Characterization of Waste Glasses with Varying Processing Temperature  

SciTech Connect (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

243

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

E-Print Network [OSTI]

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

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

1983-01-01T23:59:59.000Z

244

Two-stage thermal/nonthermal waste treatment process  

SciTech Connect (OSTI)

An innovative waste treatment technology is being developed in Los Alamos to address the destruction of hazardous organic wastes. The technology described in this report uses two stages: a packed bed reactor (PBR) in the first stage to volatilize and/or combust liquid organics and a silent discharge plasma (SDP) reactor to remove entrained hazardous compounds in the off-gas to even lower levels. We have constructed pre-pilot-scale PBR-SDP apparatus and tested the two stages separately and in combined modes. These tests are described in the report.

Rosocha, L.A.; Anderson, G.K.; Coogan, J.J.; Kang, M.; Tennant, R.A.; Wantuck, P.J.

1993-05-01T23:59:59.000Z

245

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

SciTech Connect (OSTI)

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

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

2012-10-19T23:59:59.000Z

246

Development of a Performance and Processing Property Acceptance Region for Cementitious Low-Level Waste Forms at Savannah River Site - 13174  

SciTech Connect (OSTI)

The Saltstone Production and Disposal Facilities (SPF and SDF) at the Savannah River Site (SRS) have been treating decontaminated salt solution, a low-level aqueous waste stream (LLW) since facility commissioning in 1990. In 2012, the Saltstone Facilities implemented a new Performance Assessment (PA) that incorporates an alternate design for the disposal facility to ensure that the performance objectives of DOE Order 435.1 and the National Defense Authorization Act (NDAA) of Fiscal Year 2005 Section 3116 are met. The PA performs long term modeling of the waste form, disposal facility, and disposal site hydrogeology to determine the transport history of radionuclides disposed in the LLW. Saltstone has been successfully used to dispose of LLW in a grout waste form for 15 years. Numerous waste form property assumptions directly impact the fate and transport modeling performed in the PA. The extent of process variability and consequence on performance properties are critical to meeting the assumptions of the PA. The SPF has ensured performance property acceptability by way of implementing control strategies that ensure the process operates within the analyzed limits of variability, but efforts continue to improve the understanding of facility performance in relation to the PA analysis. A similar understanding of the impact of variability on processing parameters is important from the standpoint of the operability of the production facility. The fresh grout slurry properties (particularly slurry rheology and the rate of hydration and structure formation) of the waste form directly impact the pressure and flow rates that can be reliably processed. It is thus equally important to quantify the impact of variability on processing parameters to ensure that the design basis assumptions for the production facility are maintained. Savannah River Remediation (SRR) has been pursuing a process that will ultimately establish a property acceptance region (PAR) to incorporate elements important to both processability and long-term performance properties. This process involves characterization of both emplaced product samples from the disposal facility and laboratory-simulated samples to demonstrate the effectiveness of the lab simulation. With that basis confirmed, a comprehensive variability study using non-radioactive simulants will define the acceptable PAR, or 'operating window' for Saltstone production and disposal. This same process will be used in the future to evaluate new waste streams for disposal or changes to the existing process flowsheet. (authors)

Staub, Aaron V. [Savannah River Remediation, Aiken, SC 29808 (United States)] [Savannah River Remediation, Aiken, SC 29808 (United States); Reigel, Marissa M. [Savannah River National Lab, Aiken, SC 29808 (United States)] [Savannah River National Lab, Aiken, SC 29808 (United States)

2013-07-01T23:59:59.000Z

247

Low-Level Liquid Waste Processing Pilot Studies Using a Vibratory Shear Enhancing Process (VSEP) for Filtration  

SciTech Connect (OSTI)

A previous EPRI study evaluated potential treatment methods for the removal of iron from BWR waste streams. Of the methods investigated, high shear filtration using the vibratory shear-enhanced process (VSEP) showed the most promise to effectively and economically remove high iron concentrations from backwash receiving tank waste. A VSEP filter uses oscillatory vibration to create high shear at the surface of the filter membrane. This high shear force significantly improves the filter's resistance to fouling thereby enabling high throughputs with very little secondary waste generation. With a VSEP filter, the waste feed stream is split into two effluents- a permeate stream with little or no suspended solids and a concentrate stream with a suspended solids concentration much higher than that of the feed stream. To evaluate the feasibility of using a VSEP concept for processing typical high iron containing BWR radwaste, a surrogate feedstream containing up to 1,700 ppm iron oxide (as Fe2O3) was used. This surrogate waste simulates radioactive waste found at Exelon's Limerick and Peach Bottom (powdered resin condensate) plants, and in Hope Creek's (deep bed condensate) radwaste systems. Testing was done using a series L (laboratory scale) VSEP unit at the manufacturer's and contractor's laboratories. These tests successfully demonstrated the VSEP capability for producing highly concentrated waste streams with totally ''recyclable'' permeate (e.g., greater than 95% recovery).

Bushart, S.; Tran, P.; Asay, R.

2002-02-25T23:59:59.000Z

248

Acceptable knowledge document for INEEL stored transuranic waste -- Rocky Flats Plant waste. Revision 2  

SciTech Connect (OSTI)

This document and supporting documentation provide a consistent, defensible, and auditable record of acceptable knowledge for waste generated at the Rocky Flats Plant which is currently in the accessible storage inventory at the Idaho National Engineering and Environmental Laboratory. The inventory consists of transuranic (TRU) waste generated from 1972 through 1989. Regulations authorize waste generators and treatment, storage, and disposal facilities to use acceptable knowledge in appropriate circumstances to make hazardous waste determinations. Acceptable knowledge includes information relating to plant history, process operations, and waste management, in addition to waste-specific data generated prior to the effective date of the RCRA regulations. This document is organized to provide the reader a comprehensive presentation of the TRU waste inventory ranging from descriptions of the historical plant operations that generated and managed the waste to specific information about the composition of each waste group. Section 2 lists the requirements that dictate and direct TRU waste characterization and authorize the use of the acceptable knowledge approach. In addition to defining the TRU waste inventory, Section 3 summarizes the historical operations, waste management, characterization, and certification activities associated with the inventory. Sections 5.0 through 26.0 describe the waste groups in the inventory including waste generation, waste packaging, and waste characterization. This document includes an expanded discussion for each waste group of potential radionuclide contaminants, in addition to other physical properties and interferences that could potentially impact radioassay systems.

NONE

1998-01-23T23:59:59.000Z

249

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

SciTech Connect (OSTI)

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

Not Available

1994-03-01T23:59:59.000Z

250

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

DOE Patents [OSTI]

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

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

1987-02-25T23:59:59.000Z

251

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

DOE Patents [OSTI]

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

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

1989-01-01T23:59:59.000Z

252

Microsoft PowerPoint - waste-gen-process.ppt  

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

OS-II A-II A-III Chemical waste areas in cleanroom A-II (piranha only) OS-I (only for IPA, Methanol and Acetone) A-I (acids only) B-I (bases only) OS-II (organic solvents...

253

Environmentally-friendly organochlorine waste processing and recycling  

E-Print Network [OSTI]

vinyl chloride monomer production more efficient, and on possible ways of industrial organochlorine, and vinyl chloride monomer. Polymerization and copolymeriza- tion of typical waste with and without the C5e and polymers. Ă? 2013 Elsevier Ltd. All rights reserved. 1. Introduction Vinyl chloride monomer (VCM) is one

Volinsky, Alex A.

254

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

255

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

256

Hanford Low-Activity Waste Processing: Demonstration of the Off-Gas Recycle Flowsheet - 13443  

SciTech Connect (OSTI)

Vitrification of Hanford Low-Activity Waste (LAW) is nominally the thermal conversion and incorporation of sodium salts and radionuclides into borosilicate glass. One key radionuclide present in LAW is technetium-99. Technetium-99 is a low energy, long-lived beta emitting radionuclide present in the waste feed in concentrations on the order of 1-10 ppm. The long half-life combined with a high solubility in groundwater results in technetium-99 having considerable impact on performance modeling (as potential release to the environment) of both the waste glass and associated secondary waste products. The current Hanford Tank Waste Treatment and Immobilization Plant (WTP) process flowsheet calls for the recycle of vitrification process off-gas condensates to maximize the portion of technetium ultimately immobilized in the waste glass. This is required as technetium acts as a semi-volatile specie, i.e. considerable loss of the radionuclide to the process off-gas stream can occur during the vitrification process. To test the process flowsheet assumptions, a prototypic off-gas system with recycle capability was added to a laboratory melter (on the order of 1/200 scale) and testing performed. Key test goals included determination of the process mass balance for technetium, a non-radioactive surrogate (rhenium), and other soluble species (sulfate, halides, etc.) which are concentrated by recycling off-gas condensates. The studies performed are the initial demonstrations of process recycle for this type of liquid-fed melter system. This paper describes the process recycle system, the waste feeds processed, and experimental results. Comparisons between data gathered using process recycle and previous single pass melter testing as well as mathematical modeling simulations are also provided. (authors)

Ramsey, William G.; Esparza, Brian P. [Washington River Protection Solutions, LLC, Richland, WA 99532 (United States)] [Washington River Protection Solutions, LLC, Richland, WA 99532 (United States)

2013-07-01T23:59:59.000Z

257

Pervaporation process and use in treating waste stream from glycol dehydrator  

DOE Patents [OSTI]

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

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

1994-01-01T23:59:59.000Z

258

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

E-Print Network [OSTI]

Towards model-based control of a steam Rankine process for engine waste heat recovery Johan Peralez steam process for exhaust gas heat recovery from a spark-ignition engine, focusing in particular results on a steam process for SI engines, [3] on generic control issues and [4] which provides a comp

Paris-Sud XI, Université de

259

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

SciTech Connect (OSTI)

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

TOMASZEWSKI, T.A.

2000-04-25T23:59:59.000Z

260

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

SciTech Connect (OSTI)

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

Crawford, T.W.

1995-09-22T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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.


261

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

SciTech Connect (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

262

Alcohol-free alkoxide process for containing nuclear waste  

DOE Patents [OSTI]

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

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

1984-01-01T23:59:59.000Z

263

Savannah River Site Salt Waste Processing Facility Technology Readiness  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartment of Energy fromCommentsRevolving STATEMENT OFSanEnergyWasteSeniorAssessment

264

A New Concept: Use of Negotiations in the Hazardous Waste Facility Permitting Process in New Mexico  

SciTech Connect (OSTI)

This paper describes a unique negotiation process leading to authorization of the U.S. Department of Energy (DOE) to manage and dispose remote-handled (RH) transuranic (TRU) mixed wastes at the Waste Isolation Pilot Plant (WIPP). The negotiation process involved multiple entities and individuals brought together under authority of the New Mexico Environment Department (NMED) to discuss and resolve technical and facility operational issues flowing from an NMED-issued hazardous waste facility Draft Permit. The novel negotiation process resulted in numerous substantive changes to the Draft Permit, which were ultimately memorialised in a 'Draft Permit as Changed'. This paper discusses various aspects of the negotiation process, including events leading to the negotiations, regulatory basis for the negotiations, negotiation participants, and benefits of the process. (authors)

Johnson, G.J. [Washington TRU Solutions, LLC, Waste Isolation Pilot Plant, New Mexico (United States); Rose, W.M. [U.S. Department of Energy, Carlsbad Field Office, Waste Isolation Pilot Plant, New Mexico (United States); Domenici, P.V.; Hollingsworth, L. [Domenici Law Firm PC, Albuquerque, New Mexico (United States)

2007-07-01T23:59:59.000Z

265

Processing results of 1,800 gallons of mercury and radioactively contaminated mixed waste rinse solution  

SciTech Connect (OSTI)

The mercury-contaminated rinse solution (INEL waste ID{number_sign} 123; File 8 waste) was successfully treated at the Idaho National Engineering Laboratory (INEL). This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 (HTRE-3) reactor shield tank. Approximately 1,800 gal of waste was generated and was placed into 33 drums. Each drum contained precipitated sludge material ranging from 1--10 in. in depth, with the average depth of about 2.5 in. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act (RCRA) limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/ml, while the average sludge contamination was about 13,800 pci/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. Because of difficulties in processing, three trials were required to reduce the mercury levels to below the RCRA limit. In the first trial, insufficient filtration of the waste allowed solid particulate produced during pH adjustment to enter into the ion exchange columns and ultimately the waste storage tank. In the second trial, the waste was filtered down to 0.1 {mu} to remove all solid mercury compounds. However, before filtration could take place, a solid mercury complex dissolved and mercury levels exceeded the RCRA limit after filtration. In the third trial, the waste was filtered through 0.3-A filters and then passed through the S-920 resin to remove the dissolved mercury. The resulting solution had mercury levels at 0.0186 mg/l and radioactivity of 0.282 pCi/ml. This solution was disposed of at the TAN warm waste pond, TAN782, TSF-10.

Thiesen, B.P.

1993-01-01T23:59:59.000Z

266

Bio-processing of solid wastes and secondary resources for metal extraction - A review  

SciTech Connect (OSTI)

Highlights: Black-Right-Pointing-Pointer Review focuses on bio-extraction of metals from solid wastes of industries and consumer goods. Black-Right-Pointing-Pointer Bio-processing of certain effluents/wastewaters with metals is also included in brief. Black-Right-Pointing-Pointer Quantity/composition of wastes are assessed, and microbes used and leaching conditions included. Black-Right-Pointing-Pointer Bio-recovery using bacteria, fungi and archaea is highlighted for resource recycling. Black-Right-Pointing-Pointer Process methodology/mechanism, R and D direction and scope of large scale use are briefly included. - Abstract: Metal containing wastes/byproducts of various industries, used consumer goods, and municipal waste are potential pollutants, if not treated properly. They may also be important secondary resources if processed in eco-friendly manner for secured supply of contained metals/materials. Bio-extraction of metals from such resources with microbes such as bacteria, fungi and archaea is being increasingly explored to meet the twin objectives of resource recycling and pollution mitigation. This review focuses on the bio-processing of solid wastes/byproducts of metallurgical and manufacturing industries, chemical/petrochemical plants, electroplating and tanning units, besides sewage sludge and fly ash of municipal incinerators, electronic wastes (e-wastes/PCBs), used batteries, etc. An assessment has been made to quantify the wastes generated and its compositions, microbes used, metal leaching efficiency etc. Processing of certain effluents and wastewaters comprising of metals is also included in brief. Future directions of research are highlighted.

Lee, Jae-chun [Mineral Resources Research Division, Korea Institute of Geoscience and Mineral Resources (KIGAM), Gwahang-no, Yuseong-gu, Daejeon 305-350 (Korea, Republic of); Pandey, Banshi Dhar, E-mail: bd_pandey@yahoo.co.uk [Mineral Resources Research Division, Korea Institute of Geoscience and Mineral Resources (KIGAM), Gwahang-no, Yuseong-gu, Daejeon 305-350 (Korea, Republic of); CSIR - National Metallurgical Laboratory, Jamshedpur 831007 (India)

2012-01-15T23:59:59.000Z

267

Road Map for Development of Crystal-Tolerant High Level Waste Glasses  

SciTech Connect (OSTI)

This road map guides the research and development for formulation and processing of crystal-tolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford waste glasses without jeopardizing melter operation by crystal accumulation in the melter or melter discharge riser. The potential applicability to the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) is also addressed in this road map.

Matyas, Josef; Vienna, John D.; Peeler, David; Fox, Kevin; Herman, Connie; Kruger, Albert A.

2014-05-31T23:59:59.000Z

268

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

E-Print Network [OSTI]

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

unknown authors

269

Waste disposal and treatment in the food processing industry. (Latest citations from the Biobusiness database). Published Search  

SciTech Connect (OSTI)

The bibliography contains citations concerning waste treatment and disposal in the food processing industry. Methods, equipment, and technology are considered. References discuss waste heat recovery and examine treatment of wastes resulting from meat and seafood processing, dairy and beverage production, and fruit and vegetable processing. The citations explore conversion of the treated waste to fertilizer and for use in animal feeds, combustion for energy production, biogas production, and composting. The recovery and recycling of usable chemicals from the food waste are also covered. Food packaging recycling is considered in a related bibliography. (Contains 250 citations and includes a subject term index and title list.)

NONE

1995-01-01T23:59:59.000Z

270

Waste disposal and treatment in the food-processing industry. (Latest citations from the Biobusiness data base). Published Search  

SciTech Connect (OSTI)

The bibliography contains citations concerning waste treatment and disposal in the food processing industry. Methods, equipment, and technology are considered. Specific areas include waste heat recovery, and food industry wastes from meat and seafood processing, dairy and beverage production, and processing of fruits and vegetables. The citations explore conversion of the treated waste to fertilizer, and uses in animal feeds, combustion for energy production, biogas production, and composting. The recovery and recycling of usable chemicals from the food waste is also covered. Food packaging recycling is considered in a related bibliography. (Contains 250 citations and includes a subject term index and title list.)

Not Available

1992-08-01T23:59:59.000Z

271

Waste disposal and treatment in the food processing industry. (Latest citations from the Biobusiness database). Published Search  

SciTech Connect (OSTI)

The bibliography contains citations concerning waste treatment and disposal in the food processing industry. Methods, equipment, and technology are considered. References discuss waste heat recovery and examine treatment of wastes resulting from meat and seafood processing, dairy and beverage production, and fruit and vegetable processing. The citations explore conversion of the treated waste to fertilizer and for use in animal feeds, combustion for energy production, biogas production, and composting. The recovery and recycling of usable chemicals from the food waste are also covered. Food packaging recycling is considered in a related bibliography. (Contains 250 citations and includes a subject term index and title list.)

Not Available

1994-02-01T23:59:59.000Z

272

Waste disposal and treatment in the food processing industry. (Latest citations from the Biobusiness database). Published Search  

SciTech Connect (OSTI)

The bibliography contains citations concerning waste treatment and disposal in the food processing industry. Methods, equipment, and technology are considered. References discuss waste heat recovery and examine treatment of wastes resulting from meat and seafood processing, dairy and beverage production, and fruit and vegetable processing. The citations explore conversion of the treated waste to fertilizer and for use in animal feeds, combustion for energy production, biogas production, and composting. The recovery and recycling of usable chemicals from the food waste are also covered. Food packaging recycling is considered in a related bibliography. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

NONE

1995-12-01T23:59:59.000Z

273

Independent Oversight Assessment, Salt Waste Processing Facility Project -  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana.ProgramJulietipDepartment ofThe full textImmobilizationWaste Treatmentof2012

274

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

SciTech Connect (OSTI)

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

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

2010-11-01T23:59:59.000Z

275

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

SciTech Connect (OSTI)

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

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

1993-12-31T23:59:59.000Z

276

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

SciTech Connect (OSTI)

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

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

2009-01-01T23:59:59.000Z

277

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

SciTech Connect (OSTI)

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

278

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

SciTech Connect (OSTI)

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

279

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

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September2-SCORECARD-01-24-13 Page 1 of 1Sandra L. BurrellSarai Salt Waste

280

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

SciTech Connect (OSTI)

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

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

2007-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" from the National Library of EnergyBeta (NLEBeta).
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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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281

Development Of Ion Chromatography Methods To Support Testing Of The Glycolic Acid Reductant Flowsheet In The Defense Waste Processing Facility  

SciTech Connect (OSTI)

Ion Chromatography (IC) is the principal analytical method used to support studies of Sludge Reciept and Adjustment Tank (SRAT) chemistry at DWPF. A series of prior analytical ''Round Robin'' (RR) studies included both supernate and sludge samples from SRAT simulant, previously reported as memos, are tabulated in this report.2,3 From these studies it was determined to standardize IC column size to 4 mm diameter, eliminating the capillary column from use. As a follow on test, the DWPF laboratory, the PSAL laboratory, and the AD laboratory participated in the current analytical RR to determine a suite of anions in SRAT simulant by IC, results also are tabulated in this report. The particular goal was to confirm the laboratories ability to measure and quantitate glycolate ion. The target was + or - 20% inter-lab agreement of the analyte averages for the RR. Each of the three laboratories analyzed a batch of 12 samples. For each laboratory, the percent relative standard deviation (%RSD) of the averages on nitrate, glycolate, and oxalate, was 10% or less. The three laboratories all met the goal of 20% relative agreement for nitrate and glycolate. For oxalate, the PSAL laboratory reported an average value that was 20% higher than the average values reported by the DWPF laboratory and the AD laboratory. Because of this wider window of agreement, it was concluded to continue the practice of an additional acid digestion for total oxalate measurement. It should also be noted that large amounts of glycolate in the SRAT samples will have an impact on detection limits of near eluting peaks, namely Fluoride and Formate. A suite of scoping experiments are presented in the report to identify and isolate other potential interlaboratory disceprancies. Specific ion chromatography inter-laboratory method conditions and differences are tabulated. Most differences were minor but there are some temperature control equipment differences that are significant leading to a recommendation of a heated jacket for analytical columns that are remoted for use in radiohoods. A suggested method improvement would be to implement column temperture control at a temperature slightly above ambient to avoid peak shifting due to temperature fluctuations. Temperature control in this manner would improve short and longer term peak retention time stability. An unknown peak was observed during the analysis of glycolic acid and SRAT simulant. The unknown peak was determined to best match diglycolic acid. The development of a method for acetate is summaraized, and no significant amount of acetate was observed in the SRAT products tested. In addition, an alternative Gas Chromatograph (GC) method for glycolate is summarized.

Wiedenman, B. J.; White, T. L.; Mahannah, R. N.; Best, D. R.; Stone, M. E.; Click, D. R.; Lambert, D. P.; Coleman, C. J.

2013-10-01T23:59:59.000Z

282

CHALLENGES WITH RETRIEVING TRANSURANIC WASTE FROM THE HANFORD BURIAL GROUNDS  

SciTech Connect (OSTI)

The U.S. DOE's Hanford Reservation produced plutonium and other nuclear materials for the nation's defense starting in World War II. The defense mission generated wastes that were either retrievably stored (i.e. retrievably stored waste) and/or disposed of in burial grounds. Challenges have emerged from retrieving suspect TRU waste including adequacy of records, radiological concerns, container integrity, industrial hygiene and safety issues, the lack of processing/treatment facilities, and the integration of regulatory requirements. All retrievably stored waste is managed as mixed waste and assumed to be TRU waste, unless documented otherwise. Mixed waste is defined as radioactive waste that contains hazardous constituents. The Atomic Energy Act governs waste with radionuclides, and the Resource Conservation and Recovery Act (RCRA) governs waste with hazardous constituents. Waste may also be governed by the Toxic Substances Control Act (TSCA), and a portion may be managed under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). In 1970, TRU waste was required to be placed in 20-year retrievable storage and segregated from other Waste. Prior to that date, segregation did not occur. Because of the changing definition of TRU over the years, and the limitations of early assay equipment, all retrievably stored waste in the burial grounds is managed as suspect TRU. Experience has shown that some of this waste will be characterized as low-level (non-TRU) waste after assay. The majority of the retrieved waste is not amenable to sampling due to waste type and/or radiological issues. Key to waste retrieval and disposition are characterization, historical investigation and research, knowledge of past handling and packaging, as well as a broad understanding and application of the regulations.

SWAN, R.J.; LAKES, M.E.

2007-08-06T23:59:59.000Z

283

Recovery of valuable chemical feedstocks from waste automotive plastics via pyrolysis processes  

SciTech Connect (OSTI)

Each year in North America over 9 million scrap vehicles are shredded to recover approximately 10 million tons of ferrous metal. The process also produces 3 million tons of waste known as automobile shredder residue (ASR) which consists of plastics, rubber, foams, textiles, glass, dirt, rust, etc. This waste is currently landfilled. In this study the authors present the results obtained in three different pyrolysis processes when ASR was used as the pyrolysis feedstock. The pyrolysis processes examined included: (1) a fast pyrolysis process, featuring rapid heat transfer and short residence times. This process produced primarily a gas stream that was rich in C{sub 1} to C{sub 3} hydrocarbons; (2) a screw kiln unit, characterized by slow heating and long residence times. This process produced a liquid stream that was high in aromatics; (3) a bench-scale autoclave reactor which, in the presence of water, produced a pyrolysis liquid containing large quantities of oxygenated hydrocarbons.

Shen, Z.; Day, M.; Cooney, D. [National Research Council Canada, Ottawa, Ontario (Canada). Inst. for Environmental Research and Technology

1995-11-01T23:59:59.000Z

284

Development of Vitrification Process and Glass Formulation for Nuclear Waste Conditioning  

SciTech Connect (OSTI)

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

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

2002-02-26T23:59:59.000Z

285

Radioanalytical Chemistry for Automated Nuclear Waste Process Monitoring  

SciTech Connect (OSTI)

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

Jay W. Grate; Timothy A. DeVol

2006-07-20T23:59:59.000Z

286

Advanced Thermoelectric Materials for Efficient Waste Heat Recovery in Process Industries  

SciTech Connect (OSTI)

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

Adam Polcyn; Moe Khaleel

2009-01-06T23:59:59.000Z

287

Ceramic stabilization of hazardous wastes: a high performance room temperature process  

SciTech Connect (OSTI)

ANL has developed a room-temperature process for converting hazardous materials to a ceramic structure. It is similar to vitrification but is achieved at low cost, similar to conventional cement stabilization. The waste constituents are both chemically stabilized and physically encapsulated, producing very low leaching levels and the potential for delisting. The process, which is pH-insensitive, is ideal for inorganic sludges and liquids, as well as mixed chemical-radioactive wastes, but can also handle significant percentages of salts and even halogenated organics. High waste loadings are possible and densification occurs,so that volumes are only slightly increased and in some cases (eg, incinerator ash) are reduced. The ceramic product has strength and weathering properties far superior to cement products.

Maloney, M.D.

1996-10-01T23:59:59.000Z

288

Geochemical factors affecting radionuclide transport through near and far fields at a Low-Level Waste Disposal Site  

SciTech Connect (OSTI)

The concentration of low-level waste (LLW) contaminants in groundwater is determined by the amount of contaminant present in the solid waste, rate of release from the waste and surrounding barriers, and a number of geochemical processes including adsorption, desorption, diffusion, precipitation, and dissolution. To accurately predict radionuclide transport through the subsurface, it is essential that the important geochemical processes affecting radionuclide transport be identified and, perhaps more importantly, accurately quantified and described in a mathematically defensible manner.

Kaplan, D.I.; Seme, R.J. [Pacific Northwest Lab., Richland, WA (United States); Piepkho, M.G. [Westinghouse Hanford Co., Richland, WA (United States)

1995-03-01T23:59:59.000Z

289

Integrated process analysis of treatment systems for mixed low level waste  

SciTech Connect (OSTI)

Selection of technologies to be developed for treatment of DOE`s mixed low level waste (MLLW) requires knowledge and understanding of the expected costs, schedules, risks, performance, and reliability of the total engineered systems that use these technologies. Thus, an integrated process analysis program was undertaken to identify the characteristics and needs of several thermal and nonthermal systems. For purposes of comparison, all systems were conceptually designed for a single facility processing the same amount of waste at the same rate. Thirty treatment systems were evaluated ranging from standard incineration to innovative thermal systems and innovative nonthermal chemical treatment. Treating 236 million pounds of waste in 20 years through a central treatment was found to be the least costly option with total life cycle cost ranging from $2.1 billion for a metal melting system to $3.9 billion for a nonthermal acid digestion system. Little cost difference exists among nonthermal systems or among thermal systems. Significant cost savings could be achieved by working towards maximum on line treatment time per year; vitrifying the final waste residue; decreasing front end characterization segregation and sizing requirements; using contaminated soil as the vitrifying agent; and delisting the final vitrified waste form from Resource Conservation and Recovery Act (RCRA) Land Disposal Restriction (LDR) requirements.

Cooley, C.R. [Dept. of Energy, Washington, DC (United States); Schwinkendorf, W.E. [Lockheed Martin Idaho Technology Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.]|[Sandia National Labs., Albuquerque, NM (United States); Bechtold, T.E. [Lockheed Martin Idaho Technology Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

1997-10-01T23:59:59.000Z

290

Hazardous-waste combustion in industrial processes: cement and lime kilns  

SciTech Connect (OSTI)

This report summarizes the results of several studies relating to hazardous-waste combustion in cement and lime kilns. The tests included in the study are four kilns tested by the U.S. Environmental Protection Agency, four kilns tested by State agencies or the kiln operator, two Canadian tests, and one Swedish test. The predominant types of wastes tested included chlorinated organic compounds, aromatic compounds, and metal-contaminated waste oil. The kiln types include lime kilns and cement kilns, which included the dry, wet, and preheated processes. Fabric filters and electrostatic precipitators (ESPs) were the pollution-control devices used in these processes, and the primary fuels included coal, coke, coal/coke, fuel oil, and natural gas/coke. The parameters examined in the report were Destruction and Removal Efficiency (DRE) of the Principal Organic Hazardous Constitutents, particulate and HCl emissions, metals, and the effect of burning hazardous waste on SO/sub 2/, NOx, and CO emissions. The primary conclusion of the study is that DRE's of 99.99% or greater can be obtained in properly-operated calcining kilns. Particulate matter can increase when chlorinated wastes are burned in a kiln equipped with an electrostatic precipitator. Those kilns equipped with fabric filters showed no change in emissions.

Mournighan, R.E.; Branscome, M.

1987-11-01T23:59:59.000Z

291

Microwave applicator for in-drum processing of radioactive waste slurry  

DOE Patents [OSTI]

A microwave applicator for processing of radioactive waste slurry uses a waveguide network which splits an input microwave of TE.sub.10 rectangular mode to TE.sub.01 circular mode. A cylindrical body has four openings, each receiving 1/4 of the power input. The waveguide network includes a plurality of splitters to effect the 1/4 divisions of power.

White, Terry L. (Oak Ridge, TN)

1994-01-01T23:59:59.000Z

292

IWater Processing and Waste Management SystemsIntegrated System Health Management 2007 Phase II  

E-Print Network [OSTI]

SBIR SBIR 44 45 IWater Processing and Waste Management SystemsIntegrated System Health Management valuable and, in some cases, critical features for Integrated System Health Management (ISHM) developersDE DP) to TRL 6 or higher. To facilitate Phase III NASA transition, the second program goal is deploying

293

Bagless transfer process and apparatus for radioactive waste confinement  

DOE Patents [OSTI]

A process and apparatus is provided for removing radioactive material from a glovebox, placing the material in a stainless steel storage vessel in communication with the glovebox, and sealing the vessel with a welded plug. The vessel is then severed along the weld, a lower half of the plug forming a closure for the vessel. The remaining welded plug half provides a seal for the remnant portion of the vessel and thereby maintains the sealed integrity of the glovebox.

Maxwell, David N. (Aiken, SC); Hones, Robert H. (Evans, GA); Rogers, M. Lane (Aiken, SC)

1998-01-01T23:59:59.000Z

294

Development of a metal hydride electrode waste treatment process  

SciTech Connect (OSTI)

Manufacturing residues of metal hydride electrodes for nickel - metal hydride batteries were chemically processed to recover the metal part and heat treated for the organic part. Chemical recovery yielded Ni-Co alloy after electrolysis of the solution and hydroxides of other metal, mainly rare earths. The organic part, pyrolyzed at 700 C, led to separation between carbon and fluorinated matter. Infrared coupling at the output of the pyrolysis furnace was used to identify the pyrolysis gases.

Bianco, J.C.; Martin, D.; Ansart, F.; Castillo, S.

1999-12-01T23:59:59.000Z

295

Bagless transfer process and apparatus for radioactive waste confinement  

DOE Patents [OSTI]

A process and apparatus are provided for removing radioactive material from a glovebox, placing the material in a stainless steel storage vessel in communication with the glovebox, and sealing the vessel with a welded plug. The vessel is then severed along the weld, a lower half of the plug forming a closure for the vessel. The remaining welded plug half provides a seal for the remnant portion of the vessel and thereby maintains the sealed integrity of the glovebox. 7 figs.

Maxwell, D.N.; Hones, R.H.; Rogers, M.L.

1998-04-14T23:59:59.000Z

296

Use of the Waste-Incidental-to-Reprocessing Citation Process at the West Valley Demonstration Project - 12250  

SciTech Connect (OSTI)

The West Valley Demonstration Project recently achieved a breakthrough in management of radioactive waste from reprocessing of spent nuclear fuel by taking advantage of lessons learned at other Department of Energy (DOE) sites in implementation of the waste-incidental-to-reprocessing citation process of DOE Manual 435.1-1, Radioactive Waste Management. This breakthrough involved a revision to the site procedure on waste-incidental to reprocessing. This procedure revision served as the basis for a determination by the DOE West Valley field office using the citation process that three secondary waste streams consisting of equipment that had once been contaminated by association with HLW are not HLW following decontamination and may be disposed of as low-level waste (LLW) or transuranic waste. These waste streams, which comprised much of the approximately 380 cubic meters of West Valley waste contaminated by association with HLW, included several vessels and certain tank farm equipment. By making use of lessons learned in use of the citation process by other DOE sites and information developed to support use of the citation process at the Hanford site and the Savannah River Site, the team developed a technical basis for showing that use of the citation process of DOE Manual 435.1-1 for the three new waste stream was appropriate and technically justified. The Waste Management Working Group of the EFCOG assisted in transferring lessons learned by drawing on experience from around the DOE complex. This process shared knowledge about effective implementation of the citation process in a manner that proved to be beneficial to the West Valley Demonstration Project and resulted in a technical basis document that could be used to determine that the three new waste streams were not HLW. (authors)

Sullivan, Dan; Suttora, Linda [U.S. Department of Energy (United States); Goldston, Sonny [EnergySolutions - USA (United States); Petras, Robert [Savannah River Remediation (United States); Rowell, Laurene [URS Corporation (United States); McNeil, Jim [Consultant - USA (United States)

2012-07-01T23:59:59.000Z

297

Chemi-microbial processing of waste tire rubber: A project overview  

SciTech Connect (OSTI)

PNL is developing a method to use thiophillic microorganisms to devulcanize (biodesulfurize) the surface of ground rubber particles, which will improve the bonding and adhesion of the ground tire rubber into the virgin tire rubber matrix. The Chemi-microbial processing approach, introduced in this paper, is targeted at alleviating the waste tire problem in an environmentally conscious manner; it may also be applied to improve asphaltic materials and rubber and polymeric wastes to facilite their recycling. This paper outlines the logic and technical methods that will be used.

Romine, R.A.; Snowden-Swan, L.

1993-12-01T23:59:59.000Z

298

Design of the Waste Receiving and Processing Module 2A Facility  

SciTech Connect (OSTI)

Westinghouse Hanford Company has determined that a facility is required for the treatment of mixed low-level waste at the Hanford Site. The mission of that facility will be to receive, process/treat, package, certify, and ship the contact-handled, mixed low-level waste that must be handled by Hanford Site to permanent disposal. Preconceptual and conceptual design studies were performed by United Engineers and Constructors, and a conceptual design report was issued. This report presents a summary of the conceptual design for a facility that will meet the mission established.

Lamberd, D.L.

1993-03-01T23:59:59.000Z

299

Waste Processing Annual Technology Development Report 2007 | Department of  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment of Energy While dryWashington'sResults |Energy Processing

300

Criteria and Processes for the Certification of Non-Radioactive Hazardous and Non-Hazardous Wastes  

SciTech Connect (OSTI)

This document details Lawrence Livermore National Laboratory's (LLNL) criteria and processes for determining if potentially volumetrically contaminated or potentially surface contaminated wastes are to be managed as material containing residual radioactivity or as non-radioactive. This document updates and replaces UCRL-AR-109662, Criteria and Procedures for the Certification of Nonradioactive Hazardous Waste (Reference 1), also known as 'The Moratorium', and follows the guidance found in the U.S. Department of Energy (DOE) document, Performance Objective for Certification of Non-Radioactive Hazardous Waste (Reference 2). The 1992 Moratorium document (UCRL-AR-109662) is three volumes and 703 pages. The first volume provides an overview of the certification process and lists the key radioanalytical methods and their associated Limits of Sensitivities. Volumes Two and Three contain supporting documents and include over 30 operating procedures, QA plans, training documents and organizational charts that describe the hazardous and radioactive waste management system in place in 1992. This current document is intended to update the previous Moratorium documents and to serve as the top-tier LLNL institutional Moratorium document. The 1992 Moratorium document was restricted to certification of Resource Conservation and Recovery Act (RCRA), State and Toxic Substances Control Act (TSCA) hazardous waste from Radioactive Material Management Areas (RMMA). This still remains the primary focus of the Moratorium; however, this document increases the scope to allow use of this methodology to certify other LLNL wastes and materials destined for off-site disposal, transfer, and re-use including non-hazardous wastes and wastes generated outside of RMMAs with the potential for DOE added radioactivity. The LLNL organization that authorizes off-site transfer/disposal of a material or waste stream is responsible for implementing the requirements of this document. The LLNL Radioactive and Hazardous Waste Management (RHWM) organization is responsible for the review and maintenance of this document. It should be noted that the DOE metal recycling moratorium is still in effect and is implemented as outlined in reference 17 when metals are being dispositioned for disposal/re-use/recycling off-site. This document follows the same methodology as described in the previously approved 1992 Moratorium document. Generator knowledge and certification are the primary means of characterization. Sampling and analysis are used when there is insufficient knowledge of a waste to determine if it contains added radioactivity. Table 1 (page 12) presents a list of LLNL's analytical methods for evaluating volumetrically contaminated waste and updates the reasonably achievable analytical-method-specific Minimum Detectable Concentrations (MDCs) for various matrices. Results from sampling and analysis are compared against the maximum MDCs for the given analytical method and the sample specific MDC to determine if the sample contains DOE added volumetric radioactivity. The evaluation of an item that has a physical form, and history of use, such that accessible surfaces may be potentially contaminated, is based on DOE Order 5400.5 (Reference 3), and its associated implementation guidance document DOE G 441.1-XX, Control and Release of Property with Residual Radioactive Material (Reference 4). The guidance document was made available for use via DOE Memorandum (Reference 5). Waste and materials containing residual radioactivity transferred off-site must meet the receiving facilities Waste Acceptance Criteria (if applicable) and be in compliance with other applicable federal or state requirements.

Dominick, J

2008-12-18T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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.


301

Hanford Tank Waste Treatment and Immobilization Plant (WTP) Waste Feed Qualification Program Development Approach - 13114  

SciTech Connect (OSTI)

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is a nuclear waste treatment facility being designed and constructed for the U.S. Department of Energy by Bechtel National, Inc. and subcontractor URS Corporation (under contract DE-AC27-01RV14136 [1]) to process and vitrify radioactive waste that is currently stored in underground tanks at the Hanford Site. A wide range of planning is in progress to prepare for safe start-up, commissioning, and operation. The waste feed qualification program is being developed to protect the WTP design, safety basis, and technical basis by assuring acceptance requirements can be met before the transfer of waste. The WTP Project has partnered with Savannah River National Laboratory to develop the waste feed qualification program. The results of waste feed qualification activities will be implemented using a batch processing methodology, and will establish an acceptable range of operator controllable parameters needed to treat the staged waste. Waste feed qualification program development is being implemented in three separate phases. Phase 1 required identification of analytical methods and gaps. This activity has been completed, and provides the foundation for a technically defensible approach for waste feed qualification. Phase 2 of the program development is in progress. The activities in this phase include the closure of analytical methodology gaps identified during Phase 1, design and fabrication of laboratory-scale test apparatus, and determination of the waste feed qualification sample volume. Phase 3 will demonstrate waste feed qualification testing in support of Cold Commissioning. (authors)

Markillie, Jeffrey R.; Arakali, Aruna V.; Benson, Peter A.; Halverson, Thomas G. [Hanford Tank Waste Treatment and Immobilization Plant Project, Richland, WA 99354 (United States)] [Hanford Tank Waste Treatment and Immobilization Plant Project, Richland, WA 99354 (United States); Adamson, Duane J.; Herman, Connie C.; Peeler, David K. [Savannah River National Laboratory, Aiken, SC 29808 (United States)] [Savannah River National Laboratory, Aiken, SC 29808 (United States)

2013-07-01T23:59:59.000Z

302

Development of the fluidized bed thermal treatment process for treating mixed waste  

SciTech Connect (OSTI)

A fluidized bed system is being developed at Rocky Flats for the treatment of mixed waste (a mixture of radioactive and chemically hazardous waste). The current program builds on experience gained in the 1970`s and 1980`s in tests with bench-scale, pilot-scale, and demonstration-scale fluidized bed systems. The system operates at low temperatures ({approx} 525--600{degree}C) which eliminates many of the disadvantages associated with high temperature thermal treatment processes. The process has shown the ability to destroy polychlorinated biphenyls (PCB`s) with 99.9999% (``six-nines``) destruction efficiency in tests monitored by the Environmental Protection Agency (EPA). The bed makes use of in situ neutralization of acidic off-gases by incorporating sodium carbonate (Na{sub 2}CO{sub 3}) in the bed media. This eliminates using wet scrubbers to treat the off-gas; these produce a high volume of secondary waste. Once in operation, it is expected that the fluidized bed process will yield up to a 40:1 reduction in the volume of the waste.

Semones, G.B.; Williams, P.M.; Stiefvater, S.P.; Mitchell, D.L.; Roecker, B.D.

1993-05-01T23:59:59.000Z

303

Documentation of acceptable knowledge for Los Alamos National Laboratory Plutonium Facility TRU waste stream  

SciTech Connect (OSTI)

Characterization of transuranic waste from the LANL Plutonium Facility for certification and transportation to WIPP includes the use of acceptable knowledge as specified in the WIPP Quality Assurance Program Plan. In accordance with a site specific procedure, documentation of acceptable knowledge for retrievably stored and currently generated transuranic waste streams is in progress at LANL. A summary overview of the TRU waste inventory is complete and documented in the Sampling Plan. This document also includes projected waste generation, facility missions, waste generation processes, flow diagrams, times, and material inputs. The second part of acceptable knowledge documentation consists of assembling more detailed acceptable knowledge information into auditable records and is expected to require several years to complete. These records for each waste stream must support final assignment of waste matrix parameters, EPA hazardous waste numbers, and radionuclide characterization. They must also include a determination whether waste streams are defense waste streams for compliance with the WIPP Land Withdrawal Act. The LANL Plutonium Facility`s mission is primarily plutonium processing in basic special nuclear material (SNM) research activities to support national defense and energy programs. It currently has about 100 processes ranging from SNM recovery from residues to development of plutonium 238 heat sources for space applications. Its challenge is to characterize and certify waste streams from such diverse and dynamic operations using acceptable knowledge. This paper reports the progress on the certification of the first of these waste streams to the WIPP WAC.

Montoya, A.J.; Gruetzmacher, K.M.; Foxx, C.L.; Rogers, P.Z.

1998-03-01T23:59:59.000Z

304

Documentation of acceptable knowledge for LANL Plutonium Facility transuranic waste streams  

SciTech Connect (OSTI)

Characterization of transuranic waste from the LANL Plutonium Facility for certification and transportation to WIPP includes the use of acceptable knowledge as specified in the WIPP Quality Assurance Program Plan. In accordance with a site-specific procedure, documentation of acceptable knowledge for retrievably stored and currently generated transuranic waste streams is in progress at LANL. A summary overview of the transuranic waste inventory is complete and documented in the Sampling Plan. This document also includes projected waste generation, facility missions, waste generation processes, flow diagrams, times, and material inputs. The second part of acceptable knowledge documentation consists of assembling more detailed acceptable knowledge information into auditable records and is expected to require several years to complete. These records for each waste stream must support final assignment of waste matrix parameters, EPA hazardous waste numbers, and radionuclide characterization. They must also include a determination whether waste streams are defense waste streams for compliance with the WIPP Land Withdrawal Act. The LANL Plutonium Facility`s mission is primarily plutonium processing in basic special nuclear material (SNM) research activities to support national defense and energy programs. It currently has about 100 processes ranging from SNM recovery from residues to development of plutonium 238 heat sources for space applications. Its challenge is to characterize and certify waste streams from such diverse and dynamic operations using acceptable knowledge. This paper reports the progress on the certification of the first of these waste streams to the WIPP WAC.

Montoya, A.J.; Gruetzmacher, K.; Foxx, C.; Rogers, P.S.Z.

1998-07-01T23:59:59.000Z

305

Process for the encapsulation and stabilization of radioactive, hazardous and mixed wastes  

DOE Patents [OSTI]

The present invention provides a method for encapsulating and stabilizing radioactive, hazardous and mixed wastes in a modified sulfur cement composition. The waste may be incinerator fly ash or bottom ash including radioactive contaminants, toxic metal salts and other wastes commonly found in refuse. The process may use glass fibers mixed into the composition to improve the tensile strength and a low concentration of anhydrous sodium sulfide to reduce toxic metal solubility. The present invention preferably includes a method for encapsulating radioactive, hazardous and mixed wastes by combining substantially anhydrous wastes, molten modified sulfur cement, preferably glass fibers, as well as anhydrous sodium sulfide or calcium hydroxide or sodium hydroxide in a heated double-planetary orbital mixer. The modified sulfur cement is preheated to about 135.degree..+-.5.degree. C., then the remaining substantially dry components are added and mixed to homogeneity. The homogeneous molten mixture is poured or extruded into a suitable mold. The mold is allowed to cool, while the mixture hardens, thereby immobilizing and encapsulating the contaminants present in the ash.

Colombo, Peter (Patchogue, NY); Kalb, Paul D. (Wading River, NY); Heiser, III, John H. (Bayport, NY)

1997-11-14T23:59:59.000Z

306

Apparatus for the 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 oaf plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter.

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

1999-03-16T23:59:59.000Z

307

Apparatus for the 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.

1999-03-16T23:59:59.000Z

308

Process for Converting Waste Glass Fiber into Value Added Products, Final Report  

SciTech Connect (OSTI)

Nature of the Event: Technology demonstration. The project successfully met all of its technical objectives. Albacem has signed an exclusive licensing agreement with Vitro Minerals Inc., a specialty minerals company, to commercialize the Albacem technology (website: www.vitrominerals.com). Location: The basic research for the project was conducted in Peoria, Illinois, and Atlanta, Georgia, with third-party laboratory verification carried out in Ontario, Canada. Pilot-scale trials (multi-ton) were conducted at a facility in South Carolina. Full-scale manufacturing facilities have been designed and are scheduled for construction by Vitro Minerals during 2006 at a location in the Georgia, North Carolina, and South Carolina tri-state area. The Technology: This technology consists of a process to eliminate solid wastes generated at glass fiber manufacturing facilities by converting them to value-added materials (VCAS Pozzolans) suitable for use in cement and concrete applications. This technology will help divert up to 250,000 tpy of discarded glass fiber manufacturing wastes into beneficial use applications in the concrete construction industry. This technology can also be used for processing glass fiber waste materials reclaimed from monofills at manufacturing facilities. The addition of take-back materials and reclamation from landfills can help supply over 500,000 tpy of glass fiber waste for processing into value added products. In the Albacem process, waste glass fiber is ground to a fine powder that effectively functions as a reactive pozzolanic admixture for use in portland ce¬ment-based building materials and products, such as concrete, mortars, terrazzo, tile, and grouts. Because the waste fiber from the glass manufacturing industry is vitreous, clean, and low in iron and alkalis, the resulting pozzolan is white in color and highly consistent in chemical composition. This white pozzolan, termed VCAS Pozzolan (for Vitreous Calcium-Alumino-Silicate). is especially suited for white concrete applications where it imparts desirable benefits such as increased long-term strength and improved long-term durability of concrete products. Two U.S. patents entitled have been issued to Albacem covering the technology. Third-party validation testing has confirmed that the pozzolanic product is an excellent, high performance material that conforms to a ASTM standards and improves the strength and durability of concrete. Currently, there are no known significant competing technologies to process glass fiber manufacturing by-products and con¬vert them into value-added products. Most glass fiber-forming and fabrication wastes continue to be disposed in landfills at significant costs and with associated negative environmental impact. It is estimated that in a typical glass fiber manufactur¬ing facility, 10-20% by weight of the processed glass material is sent for dis¬posal to a landfill. Today, supplementary ce¬menting materials or mineral admixtures are key to achieving strong and durable concrete. Recovered materials such as coal fly ash, ground granulated blast furnace slag and silica fume are widely accepted and used in concrete all over the world, espe¬cially in the construction of “high performance” structures such as massive dams, bridges, subway tunnels, etc. These min¬eral admixtures are not suitable for white concrete and light-colored architectural concrete applications. Converting waste glass fibers into a high performance white pozzolan would allow white concrete producers to gain from the same durability benefits currently realized by gray concrete producers. Description of the Benefit: Albacem’s technology will enable the glass fiber industry to eliminate nearly 100% of its glass fiber produc¬tion waste streams by converting them into viable value-added products. With this technology, the glass industry can prevent the landfilling of about 250,000 tons of waste glass fiber annually. Glass manufacturers will realize improved production efficiency by reducing process costs through the elimination of solid was

Hemmings, Raymond T.

2005-12-31T23:59:59.000Z

309

Waste receiving and processing facility module 1, detailed design report  

SciTech Connect (OSTI)

WRAP 1 baseline documents which guided the technical development of the Title design included: (a) A/E Statement of Work (SOW) Revision 4C: This DOE-RL contractual document specified the workscope, deliverables, schedule, method of performance and reference criteria for the Title design preparation. (b) Functional Design Criteria (FDC) Revision 1: This DOE-RL technical criteria document specified the overall operational criteria for the facility. The document was a Revision 0 at the beginning of the design and advanced to Revision 1 during the tenure of the Title design. (c) Supplemental Design Requirements Document (SDRD) Revision 3: This baseline criteria document prepared by WHC for DOE-RL augments the FDC by providing further definition of the process, operational safety, and facility requirements to the A/E for guidance in preparing the design. The document was at a very preliminary stage at the onset of Title design and was revised in concert with the results of the engineering studies that were performed to resolve the numerous technical issues that the project faced when Title I was initiated, as well as, by requirements established during the course of the Title II design.

Not Available

1993-10-01T23:59:59.000Z

310

The production of chemicals from food processing wastes using a novel fermenter separator. Annual progress report, January 1993--March 1994  

SciTech Connect (OSTI)

The basic objective of this project is to convert waste streams from the food processing industry to usable fuels and chemicals using novel bioreactors. These bioreactors should allow economical utilization of waste (whey, waste sugars, waste starch, bottling wastes, candy wastes, molasses, and cellulosic wastes) by the production of ethanol, acetone/butanol, organic acids (acetic, lactic, and gluconic), yeast diacetyl flavor, and antifungal compounds. Continuous processes incorporating various processing improvements such as simultaneous product separation and immobilized cells are being developed to allow commercial scale utilization of waste stream. The production of ethanol by a continuous reactor-separator is the process closest to commercialization with a 7,500 liter pilot plant presently sited at an Iowa site to convert whey lactose to ethanol. Accomplishments during 1993 include installation and start-up of a 7,500 liter ICRS for ethanol production at an industry site in Iowa; Donation and installation of a 200 liter yeast pilot Plant to the project from Kenyon Enterprises; Modeling and testing of a low energy system for recovery of ethanol from vapor is using a solvent absorption/extractive distillation system; Simultaneous saccharification/fermentation of raw corn grits and starch in a stirred reactor/separator; Testing of the ability of `koji` process to ferment raw corn grits in a `no-cook` process.

Dale, M.C.; Venkatesh, K.V.; Choi, H.; Salicetti-Piazza, L.; Borgos-Rubio, N.; Okos, M.R.; Wankat, P.C.

1994-03-15T23:59:59.000Z

311

Building Stakeholder Trust: Defensible Government Decisions - 13110  

SciTech Connect (OSTI)

Administrative decisions must be grounded in reasonable expectations, founded on sound principles, and bounded by societal norms. Without these first principles, attaining and retaining public trust is a Herculean task. Decisions made by governmental administrators must be both transparent and defensible: without the former the agency will lose the public's trust and support (possibly prompting a legal challenge to the decision) and without the latter the decision may fail to withstand judicial scrutiny. This presentation and accompanying paper delves into the process by which governmental decisions can achieve both defensibility and openness through building stakeholder trust with transparency. Achieving and maintaining stakeholder trust is crucial, especially in the environs of nuclear waste management. Proving confidence, stability, and security to the surrounding citizenry as well as those throughout the country is the goal of governmental nuclear waste remediation. Guiding administrative decision-making processes and maintaining a broad bandwidth of communication are of incalculable importance to all those charged with serving the public, but are especially essential to those whose decisional impacts will be felt for millennia. A strong, clear, and concise administrative record documenting discrete decisions and overarching policy choices is the strongest defense to a decisional challenge. However, this can be accomplished using transparency as the fundamental building block. This documentation allows the decision-makers to demonstrate the synthesis of legal and technical challenges and fortifies the ground from which challenges will be defended when necessary. Further, administrative actions which capture the public's interest and captivate that interest throughout the process will result in a better-informed, more deeply-involved, and more heavily-invested group of interested parties. Management of information, involvement, and investment on the front-end of the process reaps rewards far more efficiently than attempts to assuage and mitigate the concerns of those parties after the fact and there are a number of tools Savannah River Remediation (SRR) has deployed that render transparency an ally in this context. The makers, applicators, and beneficiaries of policies and decisions will all benefit from strong administrative records which document decisional choices in an open and transparent manner and from timely, up-front management of concerns of interested parties. The strongest defense to decisional challenges is an ability to demonstrate the basis of the decision and the reason(s) that the decision was chosen over other alternatives. Providing a sound basis for defending challenges rather than avoiding or fighting over them allows the deciding entity the greatest opportunity to produce value for its customer. Often, a transparent process that invites public participation and is open for public review and comment will thwart challenge genesis. An entity that has to devote resources to defending its choices obviously cannot utilize those resources to further its mission. (authors)

Franklin, Victor A. [Savannah River Remediation LLC, Bldg. 705-1C, Aiken, SC 29808 (United States)] [Savannah River Remediation LLC, Bldg. 705-1C, Aiken, SC 29808 (United States)

2013-07-01T23:59:59.000Z

312

Activated-sludge process: Waste treatment. January 1985-July 1989 (Citations from the Biobusiness data base). Report for January 1985-July 1989  

SciTech Connect (OSTI)

This bibliography contains citations concerning the use of the activated sludge process in waste and waste-water treatment. Biochemistry of the activated sludge process, the effects of various pollutants on process activity, effects of environmental variables such as oxygen and water levels, nutrient requirements of microorganisms employed in activated-sludge processes, and the application of the process to specific wastes such as pharmaceuticals, halocarbons, metallic wastes, dairy wastes, coke-plant waste waters, and petrochemical effluents are among the topics discussed. Operating experiences at large plants are included. (Contains 154 citations fully indexed and including a title list.)

Not Available

1989-08-01T23:59:59.000Z

313

UNITED STATES DEPARTMENT OF ENERGY WASTE PROCESSING ANNUAL TECHNOLOGY DEVELOPMENT REPORT 2007  

SciTech Connect (OSTI)

The Office of Environmental Management's (EM) Roadmap, U.S. Department of Energy--Office of Environmental Management Engineering & Technology Roadmap (Roadmap), defines the Department's intent to reduce the technical risk and uncertainty in its cleanup programs. The unique nature of many of the remaining facilities will require a strong and responsive engineering and technology program to improve worker and public safety, and reduce costs and environmental impacts while completing the cleanup program. The technical risks and uncertainties associated with cleanup program were identified through: (1) project risk assessments, (2) programmatic external technical reviews and technology readiness assessments, and (3) direct site input. In order to address these needs, the technical risks and uncertainties were compiled and divided into the program areas of: Waste Processing, Groundwater and Soil Remediation, and Deactivation and Decommissioning (D&D). Strategic initiatives were then developed within each program area to address the technical risks and uncertainties in that program area. These strategic initiatives were subsequently incorporated into the Roadmap, where they form the strategic framework of the EM Engineering & Technology Program. The EM-21 Multi-Year Program Plan (MYPP) supports the goals and objectives of the Roadmap by providing direction for technology enhancement, development, and demonstrations that will lead to a reduction of technical uncertainties in EM waste processing activities. The current MYPP summarizes the strategic initiatives and the scope of the activities within each initiative that are proposed for the next five years (FY2008-2012) to improve safety and reduce costs and environmental impacts associated with waste processing; authorized budget levels will impact how much of the scope of activities can be executed, on a year-to-year basis. As a result of the importance of reducing technical risk and uncertainty in the EM Waste Processing programs, EM-21 has focused considerable effort on identifying the key areas of risk in the Waste Processing programs. The resulting summary of technical risks and needs was captured in the Roadmap. The Roadmap identifies key Waste Processing initiative areas where technology development work should be focused. These areas are listed below, along with the Work Breakdown Structure (WBS) designation given to each initiative area. The WBS designations will be used throughout this document.

Bush, S

2008-08-12T23:59:59.000Z

314

History of Hanford Site Defense Production (Brief)  

SciTech Connect (OSTI)

This paper acquaints the audience with the history of the Hanford Site, America's first full-scale defense plutonium production site. The paper includes the founding and basic operating history of the Hanford Site, including World War II construction and operations, three major postwar expansions (1947-55), the peak years of production (1956-63), production phase downs (1964-the present), a brief production spurt from 1984-86, the end of the Cold War, and the beginning of the waste cleanup mission. The paper also delineates historical waste practices and policies as they changed over the years at the Hanford Site, past efforts to chemically treat, ''fractionate,'' and/or immobilize Hanford's wastes, and resulting major waste legacies that remain today. This paper presents original, primary-source research into the waste history of the Hanford Site. Finally, the paper places the current Hanford Site waste remediation endeavors in the broad context of American and world history.

GERBER, M S

2001-02-01T23:59:59.000Z

315

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

SciTech Connect (OSTI)

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

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

1993-12-31T23:59:59.000Z

316

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

317

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.

Watson, John L. (Rolla, MO)

1990-01-01T23:59:59.000Z

318

Acceptable Knowledge Summary Report for Waste Stream: SR-T001-221F-HET/Drums  

SciTech Connect (OSTI)

Since beginning operations in 1954, the Department of Energy's Savannah River Site FB-Line conducted atomic energy defense activities consistent with the listing in Section 10101(3) of the Nuclear Waste Policy Act of 1982. The facility mission was to process and convert dilute plutonium solution into highly purified weapons grade plutonium metal. As a result of various activities conducted in support of the mission (e.g., operation, maintenance, repair, clean up, and facility modifications), the facility generated transuranic waste. This document, along with referenced supporting documents, provides a defensible and auditable record of acceptable knowledge for one of the waste streams from the FB-Line. The waste was packaged in 55-gallon drums, then shipped to the transuranic waste storage facility in ''E'' area of the Savannah River Site. This acceptable knowledge report includes information relating to the facility's history, configuration,equipment, process operations, and waste management practices.

Lunsford, G.F.

1999-08-23T23:59:59.000Z

319

Process Options Description for Vitrification Flowsheet Model of INEEL Sodium Bearing Waste  

SciTech Connect (OSTI)

The purpose of this document is to provide the technical information to Savannah River Site (SRS) personnel that is required for the development of a basic steady-state process simulation of the vitrification treatment train of sodium bearing waste (SBW) at Idaho National Engineering and nvironmental Laboratory (INEEL). INEEL considers simulation to have an important role in the integration/optimization of treatment process trains for the High Level Waste (HLW) Program. This project involves a joint Technical Task Plan (TTP ID77WT31, Subtask C) between SRS and INEEL. The work scope of simulation is different at the two sites. This document addresses only the treatment of SBW at INEEL. The simulation model(s) is to be built by SRS for INEEL in FY-2001.

Nichols, Todd Travis; Taylor, Dean Dalton; Lauerhass, Lance; Barnes, Charles Marshall

2001-02-01T23:59:59.000Z

320

Waste Receiving and Processing Facility Module 2A: Advanced Conceptual Design Report. Volume 1  

SciTech Connect (OSTI)

This ACDR was performed following completed of the Conceptual Design Report in July 1992; the work encompassed August 1992 to January 1994. Mission of the WRAP Module 2A facility is to receive, process, package, certify, and ship for permanent burial at the Hanford site disposal facilities the Category 1 and 3 contact handled low-level radioactive mixed wastes that are currently in retrievable storage at Hanford and are forecast to be generated over the next 30 years by Hanford, and waste to be shipped to Hanford from about DOE sites. This volume provides an introduction to the ACDR process and the scope of the task along with a project summary of the facility, treatment technologies, cost, and schedule. Major areas of departure from the CDR are highlighted. Descriptions of the facility layout and operations are included.

Not Available

1994-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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.


321

Waste Heat Reduction and Recovery for Improving Furnace Efficiency, Productivity and Emissions Performance: A BestPractices Process Heating Technical Brief  

SciTech Connect (OSTI)

This technical brief is a guide to help plant operators reduce waste heat losses associated with process heating equipment.

Not Available

2004-11-01T23:59:59.000Z

322

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

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartment of Energy fromCommentsRevolving STATEMENT OFSanEnergyWaste Processing

323

Microwave applicator for in-drum processing of radioactive waste slurry  

DOE Patents [OSTI]

A microwave applicator for processing of radioactive waste slurry uses a waveguide network which splits an input microwave of TE[sub 10] rectangular mode to TE[sub 01] circular mode. A cylindrical body has four openings, each receiving 1/4 of the power input. The waveguide network includes a plurality of splitters to effect the 1/4 divisions of power. 4 figures.

White, T.L.

1994-06-28T23:59:59.000Z

324

Electrodialysis-based separation process for salt recovery and recycling from waste water  

DOE Patents [OSTI]

A method for recovering salt from a process stream containing organic contaminants is provided, comprising directing the waste stream to a desalting electrodialysis unit so as to create a concentrated and purified salt permeate and an organic contaminants-containing stream, and contacting said concentrated salt permeate to a water-splitting electrodialysis unit so as to convert the salt to its corresponding base and acid. 6 figs.

Tsai, S.P.

1997-07-08T23:59:59.000Z

325

Electrodialysis-based separation process for salt recovery and recycling from waste water  

DOE Patents [OSTI]

A method for recovering salt from a process stream containing organic contaminants is provided, comprising directing the waste stream to a desalting electrodialysis unit so as to create a concentrated and purified salt permeate and an organic contaminants containing stream, and contacting said concentrated salt permeate to a water-splitting electrodialysis unit so as to convert the salt to its corresponding base and acid.

Tsai, Shih-Perng (Naperville, IL)

1997-01-01T23:59:59.000Z

326

Proceedings of the tenth annual DOE low-level waste management conference: Session 3: Disposal technology and facility development  

SciTech Connect (OSTI)

This document contains ten papers on various aspects of low-level radioactive waste management. Topics include: design and construction of a facility; alternatives to shallow land burial; the fate of tritium and carbon 14 released to the environment; defense waste management; engineered sorbent barriers; remedial action status report; and the disposal of mixed waste in Texas. Individual papers were processed separately for the data base. (TEM)

Not Available

1988-12-01T23:59:59.000Z

327

DWPF waste form compliance plan (Draft Revision)  

SciTech Connect (OSTI)

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

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

1991-12-31T23:59:59.000Z

328

DWPF waste form compliance plan (Draft Revision)  

SciTech Connect (OSTI)

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

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

1991-01-01T23:59:59.000Z

329

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

SciTech Connect (OSTI)

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

Brown, A.

2014-04-27T23:59:59.000Z

330

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

SciTech Connect (OSTI)

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

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

2007-07-01T23:59:59.000Z

331

Repository disposal requirements for commercial transuranic wastes (generated without reprocessing)  

SciTech Connect (OSTI)

This report forms a preliminary planning basis for disposal of commercial transuranic (TRU) wastes in a geologic repository. Because of the unlikely prospects for commercial spent nuclear fuel reprocessing in the near-term, this report focuses on TRU wastes generated in a once-through nuclear fuel cycle. The four main objectives of this study were to: develop estimates of the current inventories, projected generation rates, and characteristics of commercial TRU wastes; develop proposed acceptance requirements for TRU wastes forms and waste canisters that ensure a safe and effective disposal system; develop certification procedures and processing requirements that ensure that TRU wastes delivered to a repository for disposal meet all applicable waste acceptance requirements; and identify alternative conceptual strategies for treatment and certification of commercial TRU first objective was accomplished through a survey of commercial producers of TRU wastes. The TRU waste acceptance and certification requirements that were developed were based on regulatory requirements, information in the literature, and from similar requirements already established for disposal of defense TRU wastes in the Waste Isolation Pilot Plant (WIPP) which were adapted, where necessary, to disposal of commercial TRU wastes. The results of the TRU waste-producer survey indicated that there were a relatively large number of producers of small quantities of TRU wastes.

Daling, P.M.; Ludwick, J.D.; Mellinger, G.B.; McKee, R.W.

1986-06-01T23:59:59.000Z

332

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

SciTech Connect (OSTI)

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

333

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

SciTech Connect (OSTI)

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

334

Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering  

DOE Patents [OSTI]

In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved.

Vijayan, Sivaraman (Deep River, CA); Wong, Chi F. (Pembroke, CA); Buckley, Leo P. (Deep River, CA)

1994-01-01T23:59:59.000Z

335

Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering  

DOE Patents [OSTI]

In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved. 1 fig.

Vijayan, S.; Wong, C.F.; Buckley, L.P.

1994-11-22T23:59:59.000Z

336

Recovery of Valuable Chlorosilane Intermediates by a Novel Waste Conversion Process, Phase IIIB (Progress)  

SciTech Connect (OSTI)

From June 1998 through September 1999, direct process residue (DPR, a waste byproduct) hydrogenolysis has been studied at a large pilot plant within Dow Corning's Carrollton, KY, facility. The system reacts filtered DPR with chlorosilane monomers at high temperature and pressure. The process routinely demonstrates DPR conversions from 59% to 89% on a monthly basis. The reaction product contains high concentrations of valuable monomers such as dimethyldichlorosilane and methyldichlorosilane. An expansion of the current unit's capacity is planned to be on-line by the end of CY2000. Furthermore, a larger DPR hydrogenolysis reactor based on these results is being designed for operation in Europe at Dow Corning's Barry, Wales, site.

Kurt E. Anderson

2000-03-31T23:59:59.000Z

337

Status of development of actinide blanket processing flowsheets for accelerator transmutation of nuclear waste  

SciTech Connect (OSTI)

An accelerator-driven subcritical nuclear system is briefly described that transmutes actinides and selected long-lived fission products. An application of this accelerator transmutation of nuclear waste (ATW) concept to spent fuel from a commercial nuclear power plant is presented as an example. The emphasis here is on a possible aqueous processing flowsheet to separate the actinides and selected long-lived fission products from the remaining fission products within the transmutation system. In the proposed system the actinides circulate through the thermal neutron flux as a slurry of oxide particles in heavy water in two loops with different average residence times: one loop for neptunium and plutonium and one for americium and curium. Material from the Np/Pu loop is processed with a short cooling time (5-10 days) because of the need to keep the total actinide inventory, low for this particular ATW application. The high radiation and thermal load from the irradiated material places severe constraints on the separation processes that can be used. The oxide particles are dissolved in nitric acid and a quarternary, ammonium anion exchanger is used to extract neptunium, plutonium, technetium, and palladium. After further cooling (about 90 days), the Am, Cm and higher actinides are extracted using a TALSPEAK-type process. The proposed operations were chosen because they have been successfully tested for processing high-level radioactive fuels or wastes in gram to kilogram quantities.

Dewey, H.J.; Jarvinen, G.D.; Marsh, S.F.; Schroeder, N.C.; Smith, B.F.; Villarreal, R.; Walker, R.B.; Yarbro, S.L.; Yates, M.A.

1993-09-01T23:59:59.000Z

338

UNCLASSIFIHED DEFENSE DOCUMENTATION CENTER  

E-Print Network [OSTI]

UNCLASSIFIHED AD 463473 DEFENSE DOCUMENTATION CENTER FOR SCIENTIFIC AND TECHNICAL INFORMATION .... John Barton Head OR Analysis Group R. H. Krolick Manager Applied Science Laboratory Prepared for the .J

Block, Marco

339

Waste Processing To Support {sup 99}Mo Production at Sandia National Laboratories  

SciTech Connect (OSTI)

As part of the Isotope Production Program at Sandia National Laboratories New Mexico (SNL/NM), procedures are being finalized for the production of {sup 99}Mo from the irradiation of {sup 235}U-coated stainless steel targets at the Technical Area (TA) V reactor and hot cell facilities. Methods have been identified and tested for the management of the non-product (waste) material as the final step in the production process. These methods were developed utilizing the waste material from a series of cold and hot tests, beginning with depleted uranium powder and culminating with a test involving an irradiated {sup 235}U target with an initial fission product inventory of approximately 18,000 Ci at the end of the irradiation cycle.

Longley, Susan; Carson, Susan; McDonald, Marion

1997-06-01T23:59:59.000Z

340

Greater-than-Class C low-level waste characterization technical review process  

SciTech Connect (OSTI)

Existing volume projections of greater-than-Class C low-level waste (GTCC LLW) vary significantly. The Department of Energy (DOE) National Low-Level Waste Management Program (NLLWMP) has undertaken activities to develop a best estimate of GTCC LLW volumes and activities for use as the planning basis. Initial information about the generation of GTCC LLW was obtained through a DOE Energy Information Administration survey. That information, combined with information from other related literature, formed the basis of a computer model, which projects potential GTCC LLW. This paper describes uncertainties in existing GTCC LLW characterization and volume projections data and describes the technical review process that is being used to assist in projections of GTCC LLW expected for storage and disposal. 8 refs., 2 tabs.

Hutchison, D.; Magleby, M.

1990-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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

The National Nuclear Laboratory's Approach to Processing Mixed Wastes and Residues - 13080  

SciTech Connect (OSTI)

The National Nuclear Laboratory (NNL) treats a wide variety of materials produced as by-products of the nuclear fuel cycle, mostly from uranium purification and fuel manufacture but also including materials from uranium enrichment and from the decommissioning of obsolete plants. In the context of this paper, treatment is defined as recovery of uranium or other activity from residues, the recycle of uranium to the fuel cycle or preparation for long term storage and the final disposal or discharge to the environment of the remainder of the material. NNL's systematic but flexible approach to residue assessment and treatment is described in this paper. The approach typically comprises up to five main phases. The benefits of a systematic approach to waste and residue assessments and processing are described in this paper with examples used to illustrate each phase of work. Benefits include early identification of processing routes or processing issues and the avoidance of investment in inappropriate and costly plant or processes. (authors)

Greenwood, Howard; Docrat, Tahera; Allinson, Sarah J.; Coppersthwaite, Duncan P.; Sultan, Ruqayyah; May, Sarah [National Nuclear Laboratory, Springfields, Preston, UK, PR4 0XJ (United Kingdom)] [National Nuclear Laboratory, Springfields, Preston, UK, PR4 0XJ (United Kingdom)

2013-07-01T23:59:59.000Z

342

Independent engineering review of the Hanford Waste Vitrification System  

SciTech Connect (OSTI)

The Hanford Waste Vitrification Plant (HWVP) was initiated in June 1987. The HWVP is an essential element of the plan to end present interim storage practices for defense wastes and to provide for permanent disposal. The project start was justified, in part, on efficient technology and design information transfer from the prototype Defense Waste Processing Facility (DWPF). Development of other serial Hanford Waste Vitrification System (HWVS) elements, such as the waste retrieval system for the double-shell tanks (DSTs), and the pretreatment system to reduce the waste volume converted into glass, also was required to accomplish permanent waste disposal. In July 1991, at the time of this review, the HWVP was in the Title 2 design phase. The objective of this technical assessment is to determine whether the status of the technology development and engineering practice is sufficient to provide reasonable assurance that the HWVP and the balance of the HWVS system will operate in an efficient and cost-effective manner. The criteria used to facilitate a judgment of potential successful operation are: vitrification of high-level radioactive waste from specified DSTs on a reasonably continuous basis; and glass produced with physical and chemical properties formally acknowledge as being acceptable for disposal in a repository for high-level radioactive waste. The criteria were proposed specifically for the Independent Engineering Review to focus that assessment effort. They are not represented as the criteria by which the Department will judge the prudence of the Project. 78 refs., 10 figs., 12 tabs.

Not Available

1991-10-01T23:59:59.000Z

343

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 walls. Air Products tasked our team to design an insert to place in the tubes of the WHB to increase flow velocity, thereby reducing fouling of the WHB. Objectives Air Products wishes that our team

Demirel, Melik C.

344

DOE underground storage tank waste remediation chemical processing hazards. Part I: Technology dictionary  

SciTech Connect (OSTI)

This document has been prepared to aid in the development of Regulating guidelines for the Privatization of Hanford underground storage tank waste remediation. The document has been prepared it two parts to facilitate their preparation. Part II is the primary focus of this effort in that it describes the technical basis for established and potential chemical processing hazards associated with Underground Storage Tank (UST) nuclear waste remediation across the DOE complex. The established hazards involve those at Sites for which Safety Analysis Reviews (SARs) have already been prepared. Potential hazards are those involving technologies currently being developed for future applications. Part I of this document outlines the scope of Part II by briefly describing the established and potential technologies. In addition to providing the scope, Part I can be used as a technical introduction and bibliography for Regulatory personnel new to the UST waste remediation, and in particular Privatization effort. Part II of this document is not intended to provide examples of a SAR Hazards Analysis, but rather provide an intelligence gathering source for Regulatory personnel who must eventually evaluate the Privatization SAR Hazards Analysis.

DeMuth, S.F.

1996-10-01T23:59:59.000Z

345

Characterization of decontamination and decommissioning wastes expected from the major processing facilities in the 200 Areas  

SciTech Connect (OSTI)

This study was intended to characterize and estimate the amounts of equipment and other materials that are candidates for removal and subsequent processing in a solid waste facility when the major processing and handling facilities in the 200 Areas of the Hanford Site are decontaminated and decommissioned. The facilities in this study were selected based on processing history and on the magnitude of the estimated decommissioning cost cited in the Surplus Facilities Program Plan; Fiscal Year 1993 (Winship and Hughes 1992). The facilities chosen for this study include B Plant (221-B), T Plant (221-T), U Plant (221-U), the Uranium Trioxide (UO{sub 3}) Plant (224-U and 224-UA), the Reduction Oxidation (REDOX) or S Plant (202-S), the Plutonium Concentration Facility for B Plant (224-B), and the Concentration Facility for the Plutonium Finishing Plant (PFP) and REDOX (233-S). This information is required to support planning activities for current and future solid waste treatment, storage, and disposal operations and facilities.

Amato, L.C.; Franklin, J.D.; Hyre, R.A.; Lowy, R.M.; Millar, J.S.; Pottmeyer, J.A. [Los Alamos Technical Associates, Kennewick, WA (United States); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (United States)

1994-08-01T23:59:59.000Z

346

Fluid bed gasification – Plasma converter process generating energy from solid waste: Experimental assessment of sulphur species  

SciTech Connect (OSTI)

Highlights: • We investigate gaseous sulphur species whilst gasifying sulphur-enriched wood pellets. • Experiments performed using a two stage fluid bed gasifier – plasma converter process. • Notable SO{sub 2} and relatively low COS levels were identified. • Oxygen-rich regions of the bed are believed to facilitate SO{sub 2}, with a delayed release. • Gas phase reducing regions above the bed would facilitate more prompt COS generation. - Abstract: Often perceived as a Cinderella material, there is growing appreciation for solid waste as a renewable content thermal process feed. Nonetheless, research on solid waste gasification and sulphur mechanisms in particular is lacking. This paper presents results from two related experiments on a novel two stage gasification process, at demonstration scale, using a sulphur-enriched wood pellet feed. Notable SO{sub 2} and relatively low COS levels (before gas cleaning) were interesting features of the trials, and not normally expected under reducing gasification conditions. Analysis suggests that localised oxygen rich regions within the fluid bed played a role in SO{sub 2}’s generation. The response of COS to sulphur in the feed was quite prompt, whereas SO{sub 2} was more delayed. It is proposed that the bed material sequestered sulphur from the feed, later aiding SO{sub 2} generation. The more reducing gas phase regions above the bed would have facilitated COS – hence its faster response. These results provide a useful insight, with further analysis on a suite of performed experiments underway, along with thermodynamic modelling.

Morrin, Shane, E-mail: shane.morrin@ucl.ac.uk [Department of Chemical Engineering, University College London, London WC1E 7JE (United Kingdom); Advanced Plasma Power, Swindon, Wiltshire SN3 4DE (United Kingdom); Lettieri, Paola, E-mail: p.lettieri@ucl.ac.uk [Department of Chemical Engineering, University College London, London WC1E 7JE (United Kingdom); Chapman, Chris, E-mail: chris.chapman@app-uk.com [Advanced Plasma Power, Swindon, Wiltshire SN3 4DE (United Kingdom); Taylor, Richard, E-mail: richard.taylor@app-uk.com [Advanced Plasma Power, Swindon, Wiltshire SN3 4DE (United Kingdom)

2014-01-15T23:59:59.000Z

347

Selection of Steady-State Process Simulation Software to Optimize Treatment of Radioactive and Hazardous Waste  

SciTech Connect (OSTI)

The process used for selecting a steady-state process simulator under conditions of high uncertainty and limited time is described. Multiple waste forms, treatment ambiguity, and the uniqueness of both the waste chemistries and alternative treatment technologies result in a large set of potential technical requirements that no commercial simulator can totally satisfy. The aim of the selection process was two-fold. First, determine the steady-state simulation software that best, albeit not completely, satisfies the requirements envelope. And second, determine if the best is good enough to justify the cost. Twelve simulators were investigated with varying degrees of scrutiny. The candidate list was narrowed to three final contenders: ASPEN Plus 10.2, PRO/II 5.11, and CHEMCAD 5.1.0. It was concluded from ''road tests'' that ASPEN Plus appears to satisfy the project's technical requirements the best and is worth acquiring. The final software decisions provide flexibility: they involve annual rather than multi-year licensing, and they include periodic re-assessment.

Nichols, T. T.; Barnes, C. M.; Lauerhass, L.; Taylor, D. D.

2001-06-01T23:59:59.000Z

348

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

SciTech Connect (OSTI)

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

349

Feed Composition for Sodium-Bearing Waste Treatment Process, Rev. 3  

SciTech Connect (OSTI)

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, Charles Marshall

2003-09-01T23:59:59.000Z

350

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

SciTech Connect (OSTI)

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

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

1993-12-31T23:59:59.000Z

351

HIGH LEVEL WASTE (HLW) VITRIFICATION EXPERIENCE IN THE US: APPLICATION OF GLASS PRODUCT/PROCESS CONTROL TO OTHERHLW AND HAZARDOUS WASTES  

SciTech Connect (OSTI)

Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. At the Savannah River Site (SRS) actual HLW tank waste has successfully been processed to stringent product and process constraints without any rework into a stable borosilicate glass waste since 1996. A unique 'feed forward' statistical process control (SPC) has been used rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. In SQC, the glass product is sampled after it is vitrified. Individual glass property models form the basis for the 'feed forward' SPC. The property models transform constraints on the melt and glass properties into constraints on the feed composition. The property models are mechanistic and depend on glass bonding/structure, thermodynamics, quasicrystalline melt species, and/or electron transfers. The mechanistic models have been validated over composition regions well outside of the regions for which they were developed because they are mechanistic. Mechanistic models allow accurate extension to radioactive and hazardous waste melts well outside the composition boundaries for which they were developed.

Jantzen, C; James Marra, J

2007-09-17T23:59:59.000Z

352

SRNL PHASE 1 ASSESSMENT OF THE WTP WASTE QUALIFICATION PROGRAM  

SciTech Connect (OSTI)

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) Project is currently transitioning its emphasis from an engineering design and construction phase toward facility completion, start-up and commissioning. With this transition, the WTP Project has initiated more detailed assessments of the requirements that must be met during the actual processing of the Hanford Site tank waste. One particular area of interest is the waste qualification program. In general, the waste qualification program involves testing and analysis to demonstrate compliance with waste acceptance criteria, determine waste processability, and demonstrate laboratory-scale unit operations to support WTP operations. The testing and analysis are driven by data quality objectives (DQO) requirements necessary for meeting waste acceptance criteria for transfer of high-level wastes from the tank farms to the WTP, and for ensuring waste processability including proper glass formulations during processing within the WTP complex. Given the successful implementation of similar waste qualification efforts at the Savannah River Site (SRS) which were based on critical technical support and guidance from the Savannah River National Laboratory (SRNL), WTP requested subject matter experts (SMEs) from SRNL to support a technology exchange with respect to waste qualification programs in which a critical review of the WTP program could be initiated and lessons learned could be shared. The technology exchange was held on July 18-20, 2011 in Richland, Washington, and was the initial step in a multi-phased approach to support development and implementation of a successful waste qualification program at the WTP. The 3-day workshop was hosted by WTP with representatives from the Tank Operations Contractor (TOC) and SRNL in attendance as well as representatives from the US DOE Office of River Protection (ORP) and the Defense Nuclear Facility Safety Board (DNFSB) Site Representative office. The purpose of the workshop was to share lessons learned and provide a technology exchange to support development of a technically defensible waste qualification program. The objective of this report is to provide a review, from SRNL's perspective, of the WTP waste qualification program as presented during the workshop. In addition to SRNL's perspective on the general approach to the waste qualification program, more detailed insight into the specific unit operations presented by WTP during the workshop is provided. This report also provides a general overview of the SRS qualification program which serves as a basis for a comparison between the two programs. Recommendations regarding specific steps are made based on the review and SRNL's lessons learned from qualification of SRS low-activity waste (LAW) and high-level waste (HLW) to support maturation of the waste qualification program leading to WTP implementation.

Peeler, D.; Hansen, E.; Herman, C.; Marra, S.; Wilmarth, B.

2012-03-06T23:59:59.000Z

353

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

SciTech Connect (OSTI)

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

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

1987-10-01T23:59:59.000Z

354

Waste disposal and treatment in the food-processing industry. March 1985-October 1989 (Citations from the Biobusiness data base). Report for March 1985-October 1989  

SciTech Connect (OSTI)

This bibliography contains citations concerning waste treatment and disposal in the food-processing industry. Methods, equipment, and technology are considered. Specific areas include waste-heat recovery, meat processing, seafood processing, dairy wastes, beverage industry, fruits and vegetables, and other food-industry wastes. Waste utilization includes animal feeds, combustion for energy production, biogas production, conversion to fertilizer, composting, and recovery and recycling of usable chemicals. Food-packaging recycling is considered in a related bibliography. (Contains 169 citations fully indexed and including a title list.)

Not Available

1989-11-01T23:59:59.000Z

355

INTERNATIONAL STUDIES OF ENHANCED WASTE LOADING AND IMPROVED MELT RATE FOR HIGH ALUMINA CONCENTRATION NUCLEAR WASTE GLASSES  

SciTech Connect (OSTI)

The goal of this study was to determine the impacts of glass compositions with high aluminum concentrations on melter performance, crystallization and chemical durability for Savannah River Site (SRS) and Hanford waste streams. Glass compositions for Hanford targeted both high aluminum concentrations in waste sludge and a high waste loading in the glass. Compositions for SRS targeted Sludge Batch 5, the next sludge batch to be processed in the Defense Waste Processing Facility (DWPF), which also has a relatively high aluminum concentration. Three frits were selected for combination with the SRS waste to evaluate their impact on melt rate. The glasses were melted in two small-scale test melters at the V. G. Khlopin Radium Institute. The results showed varying degrees of spinel formation in each of the glasses. Some improvements in melt rate were made by tailoring the frit composition for the SRS feeds. All of the Hanford and SRS compositions had acceptable chemical durability.

Fox, K; David Peeler, D; James Marra, J

2008-09-11T23:59:59.000Z

356

Study on LiCl waste salt treatment process by layer melt crystallization  

SciTech Connect (OSTI)

Layer melt crystallization operated in a static mode has been applied to separate Group I and II chlorides from surrogate LiCl waste salt. The effects of operating conditions such as crystal growing rate(or flux) and initial impurity concentration on separation (or concentration) of cesium, strontium and barium involved in a LiCl melts were analyzed. In a layer crystallization process, separation was impaired by occlusion of impurities and by residual melt adhering to LiCl crystal after at the end of the process. The crystal growth rate strongly affects the crystal structure, therefore the separation efficiency, while the effect of the initial Cs and Sr concentration in LiCl molten salt was nearly negligible. (authors)

Cho, Yung-Zun; Lee, Tae-Kyo; Choi, Jung-Hoon; Eun, Hee-Chul; Park, Hwan-Seo; Kim, In-Tae; Park, Geun-Il [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2013-07-01T23:59:59.000Z

357

The Mixed Waste Management Facility: Technology selection and implementation plan, Part 2, Support processes  

SciTech Connect (OSTI)

The purpose of this document is to establish the foundation for the selection and implementation of technologies to be demonstrated in the Mixed Waste Management Facility, and to select the technologies for initial pilot-scale demonstration. Criteria are defined for judging demonstration technologies, and the framework for future technology selection is established. On the basis of these criteria, an initial suite of technologies was chosen, and the demonstration implementation scheme was developed. Part 1, previously released, addresses the selection of the primary processes. Part II addresses process support systems that are considered ``demonstration technologies.`` Other support technologies, e.g., facility off-gas, receiving and shipping, and water treatment, while part of the integrated demonstration, use best available commercial equipment and are not selected against the demonstration technology criteria.

Streit, R.D.; Couture, S.A.

1995-03-01T23:59:59.000Z

358

Evaluation of gasification and novel thermal processes for the treatment of municipal solid waste  

SciTech Connect (OSTI)

This report identifies seven developers whose gasification technologies can be used to treat the organic constituents of municipal solid waste: Energy Products of Idaho; TPS Termiska Processor AB; Proler International Corporation; Thermoselect Inc.; Battelle; Pedco Incorporated; and ThermoChem, Incorporated. Their processes recover heat directly, produce a fuel product, or produce a feedstock for chemical processes. The technologies are on the brink of commercial availability. This report evaluates, for each technology, several kinds of issues. Technical considerations were material balance, energy balance, plant thermal efficiency, and effect of feedstock contaminants. Environmental considerations were the regulatory context, and such things as composition, mass rate, and treatability of pollutants. Business issues were related to likelihood of commercialization. Finally, cost and economic issues such as capital and operating costs, and the refuse-derived fuel preparation and energy conversion costs, were considered. The final section of the report reviews and summarizes the information gathered during the study.

Niessen, W.R.; Marks, C.H.; Sommerlad, R.E. [Camp Dresser and McKee, Inc., Cambridge, MA (United States)] [Camp Dresser and McKee, Inc., Cambridge, MA (United States)

1996-08-01T23:59:59.000Z

359

DEVELOPMENT OF A KINETIC MODEL OF BOEHMITE DISSOLUTION IN CAUSTIC SOLUTIONS APPLIED TO OPTIMIZE HANFORD WASTE PROCESSING  

SciTech Connect (OSTI)

Boehmite (e.g., aluminum oxyhydroxide) is a major non-radioactive component in Hanford and Savannah River nuclear tank waste sludge. Boehmite dissolution from sludge using caustic at elevated temperatures is being planned at Hanford to minimize the mass of material disposed of as high-level waste (HLW) during operation of the Waste Treatment Plant (WTP). To more thoroughly understand the chemistry of this dissolution process, we have developed an empirical kinetic model for aluminate production due to boehmite dissolution. Application of this model to Hanford tank wastes would allow predictability and optimization of the caustic leaching of aluminum solids, potentially yielding significant improvements to overall processing time, disposal cost, and schedule. This report presents an empirical kinetic model that can be used to estimate the aluminate production from the leaching of boehmite in Hanford waste as a function of the following parameters: (1) hydroxide concentration; (2) temperature; (3) specific surface area of boehmite; (4) initial soluble aluminate plus gibbsite present in waste; (5) concentration of boehmite in the waste; and (6) (pre-fit) Arrhenius kinetic parameters. The model was fit to laboratory, non-radioactive (e.g. 'simulant boehmite') leaching results, providing best-fit values of the Arrhenius A-factor, A, and apparent activation energy, E{sub A}, of A = 5.0 x 10{sup 12} hour{sup -1} and E{sub A} = 90 kJ/mole. These parameters were then used to predict boehmite leaching behavior observed in previously reported actual waste leaching studies. Acceptable aluminate versus leaching time profiles were predicted for waste leaching data from both Hanford and Savannah River site studies.

DISSELKAMP RS

2011-01-06T23:59:59.000Z

360

Understanding radioactive waste  

SciTech Connect (OSTI)

This document contains information on all aspects of radioactive wastes. Facts are presented about radioactive wastes simply, clearly and in an unbiased manner which makes the information readily accessible to the interested public. The contents are as follows: questions and concerns about wastes; atoms and chemistry; radioactivity; kinds of radiation; biological effects of radiation; radiation standards and protection; fission and fission products; the Manhattan Project; defense and development; uses of isotopes and radiation; classification of wastes; spent fuels from nuclear reactors; storage of spent fuel; reprocessing, recycling, and resources; uranium mill tailings; low-level wastes; transportation; methods of handling high-level nuclear wastes; project salt vault; multiple barrier approach; research on waste isolation; legal requiremnts; the national waste management program; societal aspects of radioactive wastes; perspectives; glossary; appendix A (scientific American articles); appendix B (reference material on wastes). (ATT)

Murray, R.L.

1981-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "defense waste processing" 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

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

SciTech Connect (OSTI)

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

Amoroso, J.; Marra, J.

2014-10-02T23:59:59.000Z

362

Consolidated waste forms: glass marbles and ceramic pellets  

SciTech Connect (OSTI)

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

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

1982-05-01T23:59:59.000Z

363

TECHNICAL ASSESSMENT OF BULK VITRIFICATION PROCESS & PRODUCT FOR TANK WASTE TREATMENT AT THE DEPARTMENT OF ENERGY HANFORD SITE  

SciTech Connect (OSTI)

At the U.S. Department of Energy (DOE) Hanford Site, the Waste Treatment Plant (WTP) is being constructed to immobilize both high-level waste (IUW) for disposal in a national repository and low-activity waste (LAW) for onsite, near-surface disposal. The schedule-controlling step for the WTP Project is vitrification of the large volume of LAW, current capacity of the WTP (as planned) would require 50 years to treat the Hanford tank waste, if the entire LAW volume were to be processed through the WTP. To reduce the time and cost for treatment of Hanford Tank Waste, and as required by the Tank Waste Remediation System Environmental Impact Statement Record of Decision and the Hanford Federal Facility Consent Agreement (Tn-Party Agreement), DOE plans to supplement the LAW treatment capacity of the WTP. Since 2002, DOE, in cooperation with the Environmental Protection Agency and State of Washington Department of Ecology has been evaluating technologies that could provide safe and effective supplemental treatment of LAW. Current efforts at Hanford are intended to provide additional information to aid a joint agency decision on which technology will be used to supplement the WTP. A Research, Development and Demonstration permit has been issued by the State of Washington to build and (for a limited time) operate a Demonstration Bulk Vitrification System (DBVS) facility to provide information for the decision on a supplemental treatment technology for up to 50% of the LAW. In the Bulk Vitrification (BV) process, LAW, soil, and glass-forming chemicals are mixed, dried, and placed in a refractory-lined box, Electric current, supplied through two graphite electrodes in the box, melts the waste feed, producing a durable glass waste-form. Although recent modifications to the process have resulted in significant improvements, there are continuing technical concerns.

SCHAUS, P.S.

2006-07-21T23:59:59.000Z

364

Waste-to-energy in the United States: Socioeconomic factors and the decision-making process  

SciTech Connect (OSTI)

Municipal solid waste (MSW) combustion with energy recovery, commonly called waste-to-energy (WTE), was adopted by many US communities during the 1980s to manage their growing quantities of MSW. Although less than one percent of all US MSW was burned to retrieve its heat energy in 1970, WTE grew to account for 16 percent of MSW in 1990, and many experts forecasted that WTE would be used to manage as much as half of all garbage by the turn of the century. However, the growth of WTE has been reduced in recent years by project cancellations. This study takes an in-depth look at the socioeconomic factors that have played a role in the decisions of communities that have considered WTE as a component of their solid waste management strategies. More specifically, a three-pronged approach is adopted to investigate (1) the relationships between a municipality`s decision to consider and accept/reject WTE and key socioeconomic parameters, (2) the potential impacts of recent changes in financial markets on the viability of WTE, and (3) the WTE decision-making process and the socioeconomic parameters that are most important in the municipality`s decision. The first two objectives are met by the collection and analysis of aggregate data on all US WTE initiatives during the 1982 to 1990 time frame. The latter objective is met by way of four in-depth case studies -- two directed at communities that have accepted WTE and two that have cancelled WTE projects.

Curlee, T.R.; Schexnayder, S.M.; Vogt, D.P.; Wolfe, A.K.; Kelsay, M.P.; Feldman, D.L. [Oak Ridge National Lab., TN (United States)

1993-10-01T23:59:59.000Z

365

The production of fuels and chemicals from food processing wastes using a novel fermenter separator  

SciTech Connect (OSTI)

During 1991, considerable progress was made on the waste utilization project. Two small Wisconsin companies have expressed an interest in promoting and developing the ICRS technology. Pilot plant sites at (1) Hopkinton, IA, for a sweet whey plant, and Beaver Dam WI, for an acid whey site have been under development siting ICRS operations. The Hopkinton, IA site is owned and operated by Permeate Refining Inc., who have built a batch ethanol plant across the street from Swiss Valley Farms cheddar cheese operations. Permeate from Swiss Valley is piped across to PRI. PRI has signed a contract to site a 300--500,000 gallon/yr to ICRS pilot plant. They feel that the lower labor, lower energy, continuous process offered by the ICRS will substantially improve their profitability. Catalytics, Inc, is involved with converting whey from a Kraft cream cheese operation to ethanol and yeast. A complete project including whey concentration, sterilization, and yeast growth has been designed for this site. Process design improvements with the ICRS focussed on ethanol recovery techniques during this year's project. A solvent absorption/extractive distillation (SAED) process has been developed which offers the capability of obtaining an anhydrous ethanol product from vapors off 3 to 9% ethanol solutions using very little energy for distillation. Work on products from waste streams was also performed. a. Diacetyl as a high value flavor compound was very successfully produced in a Stirred Tank Reactor w/Separation. b. Yeast production from secondary carbohydrates in the whey, lactic acid, and glycerol was studied. c. Lactic acid production from cellulose and lactose studies continued. d. Production of anti-fungal reagents by immobilized plant cells; Gossypol has antifungal properties and is produced by G. arboretum.

Dale, M.C.; Venkatesh, K.V.; Choi, Hojoon; Moelhman, M.; Saliceti, L.; Okos, M.R.; Wankat, P.C.

1991-12-01T23:59:59.000Z

366

Sandia National Laboratories: Defense Waste Management Programs  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared0EnergySandia InvolvesDOE-BER NASA

367

Technical assessment of processes to enable recycling of low-level contaminated metal waste  

SciTech Connect (OSTI)

Accumulations of metal waste exhibiting low levels of radioactivity (LLCMW) have become a national burden, both financially and environmentally. Much of this metal could be considered as a resource. The Department of Energy was assigned the task of inventorying and classifying LLCMW, identifying potential applications, and applying and/or developing the technology necessary to enable recycling. One application for recycled LLCMW is high-quality canisters for permanent repository storage of high-level waste (HLW). As many as 80,000 canisters will be needed by 2035. Much of the technology needed to decontaminate LLCMW has already been developed, but no integrated process has been described, even on a pilot scale, for recycling LLCMW into HLW canisters. This report reviews practices for removal of radionuclides and for producing low carbon stainless steel. Contaminants that readily form oxides may be reduced to below de minimis levels and combined with a slag. Most of the radioactivity remaining in the ingot is concentrated in the inclusions. Radionuclides that chemically resemble the elements that comprise stainless steel can not be removed effectively. Slag compositions, current melting practices, and canister fabrication techniques were reviewed.

Reimann, G.A.

1991-10-01T23:59:59.000Z

368

Process for removing thorium and recovering vanadium from titanium chlorinator waste  

DOE Patents [OSTI]

A process for removal of thorium from titanium chlorinator waste comprising: (a) leaching an anhydrous titanium chlorinator waste in water or dilute hydrochloric acid solution and filtering to separate insoluble minerals and coke fractions from soluble metal chlorides; (b) beneficiating the insoluble fractions from step (a) on shaking tables to recover recyclable or otherwise useful TiO.sub.2 minerals and coke; and (c) treating filtrate from step (a) with reagents to precipitate and remove thorium by filtration along with acid metals of Ti, Zr, Nb, and Ta by the addition of the filtrate (a), a base and a precipitant to a boiling slurry of reaction products (d); treating filtrate from step (c) with reagents to precipitate and recover an iron vanadate product by the addition of the filtrate (c), a base and an oxidizing agent to a boiling slurry of reaction products; and (e) treating filtrate from step (d) to remove any remaining cations except Na by addition of Na.sub.2 CO.sub.3 and boiling.

Olsen, Richard S. (Albany, OR); Banks, John T. (Corvallis, OR)

1996-01-01T23:59:59.000Z

369

Quasicrystalline Approach to Prediting the Spinel-Nepheline Liquidus: Application to Nuclear Waste Glass Processing  

SciTech Connect (OSTI)

The crystal-melt equilibria in complex fifteen component melts are modeled based on quasicrystalline concepts. A pseudobinary phase diagram between acmite (which melts incongruently to a transition metal ferrite spinel) and nepheline is defined. The pseudobinary lies within the Al{sub 2}O{sub 3}-Fe{sub 2}O{sub 3}-Na{sub 2}O-SiO{sub 2} quaternary system that defines the crystallization of basalt glass melts. The pseudobinary provides the partitioning of species between the melt and the primary liquidus phases. The medium range order of the melt and the melt-crystal exchange equilibria are defined based on a constrained mathematical treatment that considers the crystallochemical coordination of the elemental species in acmite and nepheline. The liquidus phases that form are shown to be governed by the melt polymerization and the octahedral site preference energies. This quasicrystalline liquidus model has been used to prevent unwanted crystallization in the world's largest high level waste (HLW) melter for the past three years while allowing >10 wt% higher waste loadings to be processed.

Jantzen, Carol

2005-10-10T23:59:59.000Z

370

Lessons learned from the EG&G consolidated hazardous waste subcontract and ESH&Q liability assessment process  

SciTech Connect (OSTI)

Hazardous waste transportation, treatment, recycling, and disposal contracts were first consolidated at the Idaho National Engineering Laboratory in 1992 by EG&G Idaho, Inc. At that time, disposition of Resource, Conservation and Recovery Act hazardous waste, Toxic Substance Control Act waste, Comprehensive Environmental Response, Compensation, and Liability Act hazardous substances and contaminated media, and recyclable hazardous materials was consolidated under five subcontracts. The wastes were generated by five different INEL M&O contractors, under the direction of three different Department of Energy field offices. The consolidated contract reduced the number of facilities handling INEL waste from 27 to 8 qualified treatment, storage, and disposal facilities, with brokers specifically prohibited. This reduced associated transportation costs, amount and cost of contractual paperwork, and environmental liability exposure. EG&G reviewed this approach and proposed a consolidated hazardous waste subcontract be formed for the major EG&G managed DOE sites: INEL, Mound, Rocky Flats, Nevada Test Site, and 10 satellite facilities. After obtaining concurrence from DOE Headquarters, this effort began in March 1992 and was completed with the award of two master task subcontracts in October and November 1993. In addition, the effort included a team to evaluate the apparent awardee`s facilities for environment, safety, health, and quality (ESH&Q) and financial liability status. This report documents the evaluation of the process used to prepare, bid, and award the EG&G consolidated hazardous waste transportation, treatment, recycling, and/or disposal subcontracts and associated ESH&Q and financial liability assessments; document the strengths and weaknesses of the process; and propose improvements that would expedite and enhance the process for other DOE installations that used the process and for the re-bid of the consolidated subcontract, scheduled for 1997.

Fix, N.J.

1995-03-01T23:59:59.000Z

371

Critique of Hanford Waste Vitrification Plant off-gas sampling requirements  

SciTech Connect (OSTI)

Off-gas sampling and monitoring activities needed to support operations safety, process control, waste form qualification, and environmental protection requirements of the Hanford Waste Vitrification Plant (HWVP) have been evaluated. The locations of necessary sampling sites have been identified on the basis of plant requirements, and the applicability of Defense Waste Processing Facility (DWPF) reference sampling equipment to these HWVP requirements has been assessed for all sampling sites. Equipment deficiencies, if present, have been described and the bases for modifications and/or alternative approaches have been developed.

Goles, R.W.

1996-03-01T23:59:59.000Z

372

Waste Receiving and Processing (WRAP) Facility Public Address System Review Findings  

SciTech Connect (OSTI)

Public address system operation at the Waste Receiving and Processing (WRAP) facility was reviewed. The review was based on an Operational Readiness Review finding that public address performance was not adequate in parts of the WRAP facility. Several improvements were made to the WRAP Public Address (PA) system to correct the deficiencies noted. Speaker gain and position was optimized. A speech processor was installed to boost intelligibility in high noise areas. Additional speakers were added to improve coverage in the work areas. The results of this evaluation indicate that further PA system enhancements are not warranted. Additional speakers cannot compensate for the high background sound and high reverberation levels found in the work areas. Recommendations to improve PA system intelligibility include minor speaker adjustments, enhanced PA announcement techniques, and the use of sound reduction and abatement techniques where economically feasible.

HUMPHRYS, K.L.

1999-11-03T23:59:59.000Z

373

Precipitation-adsorption process for the decontamination of nuclear waste supernates  

DOE Patents [OSTI]

High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

Lee, Lien-Mow (North Augusta, SC); Kilpatrick, Lester L. (Aiken, SC)

1984-01-01T23:59:59.000Z

374

Precipitation-adsorption process for the decontamination of nuclear waste supernates  

DOE Patents [OSTI]

High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

Lee, L.M.; Kilpatrick, L.L.

1982-05-19T23:59:59.000Z

375

Interface control document between PUREX/UO{sub 3} Plant Transition and Solid Waste Disposal Division  

SciTech Connect (OSTI)

This interface control document (ICD) between PUREX/UO{sub 3} Plant Transition (PPT) and Solid Waste Disposal Division (SWD) establishes at a top level the functional responsibilities of each division where interfaces exist between the two divisions. Since the PUREX Transition and Solid Waste Disposal divisions operate autonomously, it is important that each division has a clear understanding of the other division`s expectations regarding these interfaces. This ICD primarily deals with solid wastes generated by the PPT. In addition to delineating functional responsibilities, the ICD includes a baseline description of those wastes that will require management as part of the interface between the divisions. The baseline description of wastes includes waste volumes and timing for use in planning the proper waste management capabilities: the primary purpose of this ICD is to ensure defensibility of expected waste stream volumes and Characteristics for future waste management facilities. Waste descriptions must be as complete as-possible to ensure adequate treatment, storage, and disposal capability will exist. The ICD also facilitates integration of existing or planned waste management capabilities of the PUREX. Transition and Solid Waste Disposal divisions. The ICD does not impact or affect the existing processes or procedures for shipping, packaging, or approval for shipping wastes by generators to the Solid Waste Division.

Duncan, D.R.

1994-06-30T23:59:59.000Z

376

CONTAMINATED PROCESS EQUIPMENT REMOVAL FOR THE D&D OF THE 232-Z CONTAMINATED WASTE RECOVERY PROCESS FACILITY AT THE PLUTONIUM FINISHING PLANT (PFP)  

SciTech Connect (OSTI)

This paper describes the unique challenges encountered and subsequent resolutions to accomplish the deactivation and decontamination of a plutonium ash contaminated building. The 232-Z Contaminated Waste Recovery Process Facility at the Plutonium Finishing Plant was used to recover plutonium from process wastes such as rags, gloves, containers and other items by incinerating the items and dissolving the resulting ash. The incineration process resulted in a light-weight plutonium ash residue that was highly mobile in air. This light-weight ash coated the incinerator's process equipment, which included gloveboxes, blowers, filters, furnaces, ducts, and filter boxes. Significant airborne contamination (over 1 million derived air concentration hours [DAC]) was found in the scrubber cell of the facility. Over 1300 grams of plutonium held up in the process equipment and attached to the walls had to be removed, packaged and disposed. This ash had to be removed before demolition of the building could take place.

HOPKINS, A.M.; MINETTE, M.J.; KLOS, D.B.

2007-01-25T23:59:59.000Z

377

Verification of the Accountability Method as a Means to Classify Radioactive Wastes Processed Using THOR Fluidized Bed Steam Reforming at the Studsvik Processing Facility in Erwin, Tennessee, USA - 13087  

SciTech Connect (OSTI)

Studsviks' Processing Facility Erwin (SPFE) has been treating Low-Level Radioactive Waste using its patented THOR process for over 13 years. Studsvik has been mixing and processing wastes of the same waste classification but different chemical and isotopic characteristics for the full extent of this period as a general matter of operations. Studsvik utilizes the accountability method to track the movement of radionuclides from acceptance of waste, through processing, and finally in the classification of waste for disposal. Recently the NRC has proposed to revise the 1995 Branch Technical Position on Concentration Averaging and Encapsulation (1995 BTP on CA) with additional clarification (draft BTP on CA). The draft BTP on CA has paved the way for large scale blending of higher activity and lower activity waste to produce a single waste for the purpose of classification. With the onset of blending in the waste treatment industry, there is concern from the public and state regulators as to the robustness of the accountability method and the ability of processors to prevent the inclusion of hot spots in waste. To address these concerns and verify the accountability method as applied by the SPFE, as well as the SPFE's ability to control waste package classification, testing of actual waste packages was performed. Testing consisted of a comprehensive dose rate survey of a container of processed waste. Separately, the waste package was modeled chemically and radiologically. Comparing the observed and theoretical data demonstrated that actual dose rates were lower than, but consistent with, modeled dose rates. Moreover, the distribution of radioactivity confirms that the SPFE can produce a radiologically homogeneous waste form. The results of the study demonstrate: 1) the accountability method as applied by the SPFE is valid and produces expected results; 2) the SPFE can produce a radiologically homogeneous waste; and 3) the SPFE can effectively control the waste package classification. (authors)

Olander, Jonathan [Studsvik Processing Facility Erwin, 151 T.C. Runnion Rd., Erwin, TN 37650 (United States)] [Studsvik Processing Facility Erwin, 151 T.C. Runnion Rd., Erwin, TN 37650 (United States); Myers, Corey [Studsvik, Inc., 5605 Glenridge Drive, Suite 705, Atlanta, GA 30342 (United States)] [Studsvik, Inc., 5605 Glenridge Drive, Suite 705, Atlanta, GA 30342 (United States)

2013-07-01T23:59:59.000Z

378

Process for removing and detoxifying cadmium from scrap metal including mixed waste  

SciTech Connect (OSTI)

Cadmium-bearing scrap from nuclear applications, such as neutron shielding and reactor control and safety rods, must usually be handled as mixed waste since it is radioactive and the cadmium in it is both leachable and highly toxic. Removing the cadmium from this scrap, and converting it to a nonleachable and minimally radioactive form, would greatly simplify disposal or recycling. A process now under development will do this by shredding the scrap; leaching it with reagents which selectively dissolve out the cadmium; reprecipitating the cadmium as its highly insoluble sulfide; then fusing the sulfide into a glassy matrix to bring its leachability below EPA limits before disposal. Alternatively, the cadmium may be recovered for reuse. A particular advantage of the process is that all reagents (except the glass frit) can easily be recovered and reused in a nearly closed cycle, minimizing the risk of radioactive release. The process does not harm common metals such as aluminum, iron and stainless steel, and is also applicable to non-nuclear cadmium-bearing scrap such as nickel-cadmium batteries.

Kronberg, J.W.

1994-07-01T23:59:59.000Z

379

Remote-handled transuranic waste study  

SciTech Connect (OSTI)

The Waste Isolation Pilot Plant (WIPP) was developed by the US Department of Energy (DOE) as a research and development facility to demonstrate the safe disposal of transuranic (TRU) radioactive wastes generated from the Nation`s defense activities. The WIPP disposal inventory will include up to 250,000 cubic feet of TRU wastes classified as remote handled (RH). The remaining inventory will include contact-handled (CH) TRU wastes, which characteristically have less specific activity (radioactivity per unit volume) than the RH-TRU wastes. The WIPP Land Withdrawal Act (LWA), Public Law 102-579, requires a study of the effect of RH-TRU waste on long-term performance. This RH-TRU Waste Study has been conducted to satisfy the requirements defined by the LWA and is considered by the DOE to be a prudent exercise in the compliance certification process of the WIPP repository. The objectives of this study include: conducting an evaluation of the impacts of RH-TRU wastes on the performance assessment (PA) of the repository to determine the effects of Rh-TRU waste as a part of the total WIPP disposal inventory; and conducting a comparison of CH-TRU and RH-TRU wastes to assess the differences and similarities for such issues as gas generation, flammability and explosiveness, solubility, and brine and geochemical interactions. This study was conducted using the data, models, computer codes, and information generated in support of long-term compliance programs, including the WIPP PA. The study is limited in scope to post-closure repository performance and includes an analysis of the issues associated with RH-TRU wastes subsequent to emplacement of these wastes at WIPP in consideration of the current baseline design. 41 refs.

NONE

1995-10-01T23:59:59.000Z

380

Initial Evaluation of Processing Methods for an Epsilon Metal Waste Form  

SciTech Connect (OSTI)

During irradiation of nuclear fuel in a reactor, the five metals, Mo, Pd, Rh, Ru, and Tc, migrate to the fuel grain boundaries and form small metal particles of an alloy known as epsilon metal ({var_epsilon}-metal). When the fuel is dissolved in a reprocessing plant, these metal particles remain behind with a residue - the undissolved solids (UDS). Some of these same metals that comprise this alloy that have not formed the alloy are dissolved into the aqueous stream. These metals limit the waste loading for a borosilicate glass that is being developed for the reprocessing wastes. Epsilon metal is being developed as a waste form for the noble metals from a number of waste streams in the aqueous reprocessing of used nuclear fuel (UNF) - (1) the {var_epsilon}-metal from the UDS, (2) soluble Tc (ion-exchanged), and (3) soluble noble metals (TRUEX raffinate). Separate immobilization of these metals has benefits other than allowing an increase in the glass waste loading. These materials are quite resistant to dissolution (corrosion) as evidenced by the fact that they survive the chemically aggressive conditions in the fuel dissolver. Remnants of {var_epsilon}-metal particles have survived in the geologically natural reactors found in Gabon, Africa, indicating that they have sufficient durability to survive for {approx} 2.5 billion years in a reducing geologic environment. Additionally, the {var_epsilon}-metal can be made without additives and incorporate sufficient foreign material (oxides) that are also present in the UDS. Although {var_epsilon}-metal is found in fuel and Gabon as small particles ({approx}10 {micro}m in diameter) and has survived intact, an ideal waste form is one in which the surface area is minimized. Therefore, the main effort in developing {var_epsilon}-metal as a waste form is to develop a process to consolidate the particles into a monolith. Individually, these metals have high melting points (2617 C for Mo to 1552 C for Pd) and the alloy is expected to have a high melting point as well, perhaps exceeding 1500 C. The purpose of the work reported here is to find a potential commercial process with which {var_epsilon}-metal plus other components of UDS can be consolidated into a solid with minimum surface area and high strength Here, we report the results from the preliminary evaluation of spark-plasma sintering (SPS), hot-isostatic pressing (HIP), and microwave sintering (MS). Since bulk {var_epsilon}-metal is not available and companies could not handle radioactive materials, we prepared mixtures of the five individual metal powders (Mo, Ru, Rh, Pd, and Re) and baddeleyite (ZrO{sub 2}) to send the vendors of SPS, HIP, and MS. The processed samples were then evaluated at the Pacific Northwest National Laboratory (PNNL) for bulk density and phase assemblage with X-ray diffraction (XRD) and phase composition with scanning electron microscopy (SEM). Physical strength was evaluated qualitatively. Results of these scoping tests showed that fully dense cermet (ceramic-metal composite) materials with up to 35 mass% of ZrO{sub 2} were produced with SPS and HIP. Bulk density of the SPS samples ranged from 87 to 98% of theoretical density, while HIP samples ranged from 96 to 100% of theoretical density. Microwave sintered samples containing ZrO{sub 2} had low densities of 55 to 60% of theoretical density. Structurally, the cermet samples showed that the individual metals alloyed in to {var_epsilon}-phase - hexagonal-close-packed (HCP) alloy (4-95 mass %), the {alpha}-phase - face-centered-cubic (FCC) alloy structure (3-86 mass %), while ZrO{sub 2} remained in the monoclinic structure of baddeleyite. Elementally, the samples appeared to have nearly uniform composition, but with some areas rich in Mo and Re, the two components with the highest melting points. The homogeneity in distribution of the elements in the alloy is significantly improved in the presence of ZrO{sub 2}. However, ZrO{sub 2} does not appear to react with the alloy, nor was Zr found in the alloy.

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

2012-06-11T23:59:59.000Z

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381

Waste Heat Recovery and Recycling in Thermal Separation Processes: Distillation, Multi-Effect Evaporation (MEE) and Crystallization Processes  

SciTech Connect (OSTI)

Evaporation and crystallization are key thermal separation processes for concentrating and purifying inorganic and organic products with energy consumption over 1,000 trillion Btu/yr. This project focused on a challenging task of recovering low-temperature latent heat that can have a paradigm shift in the way thermal process units will be designed and operated to achieve high-energy efficiency and significantly reduce the carbon footprint as well as water footprint. Moreover, this project has evaluated the technical merits of waste-heat powered thermal heat pumps for recovery of latent heat from distillation, multi-effect evaporation (MEE), and crystallization processes and recycling into the process. The Project Team has estimated the potential energy, economics and environmental benefits with the focus on reduction in CO2 emissions that can be realized by 2020, assuming successful development and commercialization of the technology being developed. Specifically, with aggressive industry-wide applications of heat recovery and recycling with absorption heat pumps, energy savings of about 26.7 trillion Btu/yr have been estimated for distillation process. The direct environmental benefits of this project are the reduced emissions of combustible products. The estimated major reduction in environmental pollutants in the distillation processes is in CO2 emission equivalent to 3.5 billion lbs/year. Energy consumption associated with water supply and treatments can vary between 1,900 kWh and 23,700 kWh per million-gallon water depending on sources of natural waters [US DOE, 2006]. Successful implementation of this technology would significantly reduce the demand for cooling-tower waters, and thereby the use and discharge of water treatment chemicals. The Project Team has also identified and characterized working fluid pairs for the moderate-temperature heat pump. For an MEE process, the two promising fluids are LiNO3+KNO3+NANO3 (53:28:19 ) and LiNO3+KNO3+NANO2(53:35:12). And for an H2O2 distillation process, the two promising fluids are Trifluoroethanol (TFE) + Triethylene Glycol Dimethyl ether (DMETEG) and Ammonia+ Water. Thermo-physical properties calculated by Aspen+ are reasonably accurate. Documentation of the installation of pilot-plants or full commercial units were not found in the literature for validating thermo-physical properties in an operating unit. Therefore, it is essential to install a pilot-scale unit to verify thermo-physical properties of working fluid pairs and validate the overall efficiency of the thermal heat pump at temperatures typical of distillation processes. For an HO2 process, the ammonia-water heat pump system is more compact and preferable than the TFE-DMETEG heat pump. The ammonia-water heat pump is therefore recommended for the H2O2 process. Based on the complex nature of the heat recovery system, we anticipated that capital costs could make investments financially unattractive where steam costs are low, especially where co-generation is involved. We believe that the enhanced heat transfer equipment has the potential to significantly improve the performance of TEE crystallizers, independent of the absorption heat-pump recovery system. Where steam costs are high, more detailed design/cost engineering will be required to verify the economic viability of the technology. Due to the long payback period estimated for the TEE open system, further studies on the TEE system are not warranted unless there are significant future improvements to heat pump technology. For the H2O2 distillation cycle heat pump waste heat recovery system, there were no significant process constraints and the estimated 5 years payback period is encouraging. We therefore recommend further developments of application of the thermal heat pump in the H2O2 distillation process with the focus on the technical and economic viability of heat exchangers equipped with the state-of-the-art enhancements. This will require additional funding for a prototype unit to validate enhanced thermal performances of heat transfer equipment, evaluat

Emmanuel A. Dada; Chandrakant B. Panchal; Luke K. Achenie; Aaron Reichl; Chris C. Thomas

2012-12-03T23:59:59.000Z

382

A low-temperature process for the denitration of Hanford single-shell tank, nitrate-based waste utilizing the nitrate to ammonia and ceramic (NAC) process  

SciTech Connect (OSTI)

Bench-top feasibility studies with Hanford single-shell tank (SST) simulants, using a new, low-temperature (50 to 60C) process for converting nitrate to ammonia and ceramic (NAC), have conclusively shown that between 85 to 99% of the nitrate can be readily converted. In this process, aluminum powders or shot can be used to convert alkaline, nitrate-based supernate to ammonia and an aluminum oxide-sodium aluminate-based solid which might function as its own waste form. The process may actually be able to utilize already contaminated aluminum scrap metal from various DOE sites to effect the conversion. The final, nearly nitrate-free ceramic-like product can be pressed and sintered like other ceramics. Based upon the starting volumes of 6.2 and 3.1 M sodium nitrate solution, volume reductions of 50 to 55% were obtained for the waste form produced, compared to an expected 35 to 50% volume increase if the Hanford supernate were grouted. Engineering data extracted from bench-top studies indicate that the process will be very economical to operate, and data were used to cost a batch, 1,200-kg NO{sub 3}/h plant for working off Hanford SST waste over 20 years. Their total process cost analysis presented in the appendix, indicates that between $2.01 to 2.66 per kilogram of nitrate converted will be required. Additionally, data on the fate of select radioelements present in solution are presented in this report as well as kinetic, operational, and control data for a number of experiments. Additionally, if the ceramic product functions as its own waste form, it too will offer other cost savings associated with having a smaller volume of waste form as well as eliminating other process steps such as grouting.

Mattus, A.J.; Lee, D.D.; Dillow, T.A.; Farr, L.L.; Loghry, S.L.; Pitt, W.W.; Gibson, M.R.

1994-12-01T23:59:59.000Z

383

Field study of disposed solid wastes from advanced coal processes. Annual technical progress report, October 1991--September 1992  

SciTech Connect (OSTI)

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

Not Available

1992-12-31T23:59:59.000Z

384

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

E-Print Network [OSTI]

1 Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source. Belanger Chair, Department of Physics #12;2 Abstract A review of the sodium cooled fast reactor........................................................................................23 1.3.5 Reactor Startup

Belanger, David P.

385

Bioremediation of a Process Waste Lagoon at a Southern Polish Oil Refinery -DoE's First Demonstration Project in Poland  

E-Print Network [OSTI]

Bioremediation of a Process Waste Lagoon at a Southern Polish Oil Refinery - DoE's First by the Czechowice Oil Refinery, located in southern Poland, has produced an estimated 120 thousand tons of acidic company thereby eliminating the contaminants while providing the refinery an additional revenue source

Hazen, Terry

386

Savannah River Site Waste Isolation Pilot Plant Disposal Program - Acceptable Knowledge Summary Report for Waste Stream: SR-T001-221-HET  

SciTech Connect (OSTI)

This document, along with referenced supporting documents provides a defensible and auditable record of acceptable knowledge for one of the waste streams from the FB-Line. This heterogeneous debris transuranic waste stream was generated after January 25, 1990 and before March 20, 1997. The waste was packaged in 55-gallon drums, then shipped to the transuranic waste storage facility in ''E'' area of the Savannah River Site. This acceptable knowledge report includes information relating to the facility's history, configuration, equipment, process operations and waste management practices. Information contained in this report was obtained from numerous sources including: facility safety basis documentation, historical document archives, generator and storage facility waste records and documents, and interviews with cognizant personnel.

Lunsford, G.F.

2001-01-24T23:59:59.000Z

387

Process Options Description for Vitrification Flowsheet Model of INEEL Sodium Bearing Waste  

SciTech Connect (OSTI)

The technical information required for the development of a basic steady-state process simulation of the vitrification treatment train of sodium bearing waste (SBW) at Idaho National Engineering and Environmental Laboratory (INEEL) is presented. The objective of the modeling effort is to provide the predictive capability required to optimize an entire treatment train and assess system-wide impacts of local changes at individual unit operations, with the aim of reducing the schedule and cost of future process/facility design efforts. All the information required a priori for engineers to construct and link unit operation modules in a commercial software simulator to represent the alternative treatment trains is presented. The information is of a mid- to high-level nature and consists of the following: (1) a description of twenty-four specific unit operations--their operating conditions and constraints, primary species and key outputs, and the initial modeling approaches that will be used in the first year of the simulation's development; (2) three potential configurations of the unit operations (trains) and their interdependencies via stream connections; and (3) representative stream compositional makeups.

Nichols, T.T.; Taylor, D.D.; Lauerhass, L.; Barnes, C.M.

2002-02-21T23:59:59.000Z

388

The use of carbon aerogel electrodes for deionizing water and treating aqueous process wastes  

SciTech Connect (OSTI)

A wide variety of ionic contaminants can be removed from aqueous solutions by electrosorption on carbon aerogel electrodes. Carbon aerogel is an ideal electrode material because of its low electrical resistivity (< 40 m{Omega}-cm), high specific surface area (400 to 1100 m{sup 2}/g), and controllable pore size distribution (< 50 nm). This approach may avoid the generation of a substantial amount of secondary waste associated with ion exchange processing. Ion exchange resins require concentrated solutions of acid, base, or salt for regeneration, whereas carbon aerogel electrodes require only electrical discharge or reverse polarization. Aqueous solutions of NaCl, NaNO{sub 3}, NH{sub 4}ClO{sub 4}, Na{sub 2}CO{sub 3}, Na{sub 2}SO{sub 4} and Na{sub 3}PO{sub 4} have been separated into concentrate and high-purity product streams. The deionization of a 100 {mu}S/cm NaCl solution with two parallel stacks of carbon aerogel electrodes in a potential-swing mode is discussed in detail. The selective removal of Cu, Zn, Cd, Pb, Cr, Mn, Co and U from a variety of process solutions and natural waters has also been demonstrated. Feasibility tests indicate that the remediation of Cr(VI)-contaminated ground water may be possible.

Farmer, J.C.; Mack, G.V.; Fix, D.V.

1996-07-01T23:59:59.000Z

389

DOE final report, phase one startup, Waste Receiving and Processing Facility (WRAP)  

SciTech Connect (OSTI)

This document is to validate that the WRAP facility is physically ready to start up phase 1, and that the managers and operators are prepared to safely manage and operate the facility when all pre-start findings have been satisfactorily corrected. The DOE Readiness Assessment (RA) team spent a week on-site at Waste Receiving and Processing Module 1 (WRAP-1) to validate the readiness for phase 1 start up of facility. The Contractor and DOE staff were exceptionally cooperative and contributed significantly to the overall success of the RA. The procedures and Conduct of Operations areas had significant discrepancies, many of which should have been found by the contractor review team. In addition the findings of the contractor review team should have led the WRAP-1 management team to correcting the root causes of the findings prior to the DOE RA team review. The findings and observations include many issues that the team believes should have been found by the contractor review and corrective actions taken. A significantly improved Operational Readiness Review (ORR) process and corrective actions of root causes must be fully implemented by the contractor prior to the performance of the contractor ORR for phase 2 operations. The pre-start findings as a result of this independent DOE Readiness Assessment are presented.

Jasen, W.G.

1998-01-07T23:59:59.000Z

390

Use of the TRUEX process for the pretreatment of neutralized cladding removal waste (NCRW) sludge -- Results of FY 1990 studies  

SciTech Connect (OSTI)

The goal of this process is to separate the transuranic elements from the bulk components so that the bulk components can be disposed of as low-level waste with only a small transuranic-containing fraction requiring geologic disposal. The pretreatment process examined here is the one indicated to be most promising in the initial studies. It involves dissolving the unwashed sludge in nitric acid and then using the TRUEX solvent extraction process to remove the transuranic elements from the bulk components of the waste. The areas identified in this work that need additional information are gradual precipitate formation as dissolved sludge solutions age, and formation of solid material when the dissolved sludge solution is contacted with the solvent used in the TRUEX process. 5 refs., 71 figs., 10 tabs.

Swanson, J.L.

1991-09-01T23:59:59.000Z

391

Phosphate ceramic process for macroencapsulation and stabilization of low-level debris wastes  

SciTech Connect (OSTI)

Across the DOE complex, large quantities of contaminated debris and irradiated lead bricks have been accumulated for disposal. Under the US Environmental Protection Agency`s Alternative Treatment Standards, the preferred method of disposal of these wastes is macroencapsulation. Chemically bonded phosphate ceramic (CBPC) is a novel binder that was developed at Argonne National Laboratory to stabilize and solidify various low-level mixed wastes. This binder is extremely strong, dense, and impervious to water. In this investigation, CBPC has been used to demonstrate macroencapsulation of various contaminated debris wastes, including cryofractured debris, lead bricks, lead-lined plastic gloves, and mercury-contaminated crushed glass. This paper describes the fabrication of the waste forms, as well as the results of various characterizations performed on the waste forms. The results show that the simple and low-cost CBPC is an excellent material system for the macroencapsulation of debris wastes.

Singh, D.; Wagh, A.S.; Tlustochowicz, M.; Jeong, S.Y. [Argonne National Lab., IL (United States). Energy Technology Div.] [Argonne National Lab., IL (United States). Energy Technology Div.

1998-12-31T23:59:59.000Z

392

Quantification of greenhouse gas emissions from waste management processes for municipalities - A comparative review focusing on Africa  

SciTech Connect (OSTI)

The amount of greenhouse gases (GHG) emitted due to waste management in the cities of developing countries is predicted to rise considerably in the near future; however, these countries have a series of problems in accounting and reporting these gases. Some of these problems are related to the status quo of waste management in the developing world and some to the lack of a coherent framework for accounting and reporting of greenhouse gases from waste at municipal level. This review summarizes and compares GHG emissions from individual waste management processes which make up a municipal waste management system, with an emphasis on developing countries and, in particular, Africa. It should be seen as a first step towards developing a more holistic GHG accounting model for municipalities. The comparison between these emissions from developed and developing countries at process level, reveals that there is agreement on the magnitude of the emissions expected from each process (generation of waste, collection and transport, disposal and recycling). The highest GHG savings are achieved through recycling, and these savings would be even higher in developing countries which rely on coal for energy production (e.g. South Africa, India and China) and where non-motorized collection and transport is used. The highest emissions are due to the methane released by dumpsit