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1

Vitrified waste option study report  

SciTech Connect

A {open_quotes}Settlement Agreement{close_quotes} between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) will be treated so that it is ready to be moved out of Idaho for disposal by a target date of 2035. This report investigates vitrification treatment of all ICPP calcine, including the existing and future HLW calcine resulting from calcining liquid Sodium-Bearing Waste (SBW). Currently, the SBW is stored in the tank farm at the ICPP. Vitrification of these wastes is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the calcined waste and casting the vitrified mass into stainless steel canisters that will be ready to be moved out of the Idaho for disposal by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory (INEEL) until they are sent to a HLW national repository. The operating period for vitrification treatment will be from 2013 through 2032; all HLW will be treated and in storage by the end of 2032.

Lopez, D.A.; Kimmitt, R.R.

1998-02-01T23:59:59.000Z

2

Vitrified underground structures  

DOE Patents (OSTI)

A method of making vitrified underground structures in which 1) the vitrification process is started underground, and 2) a thickness dimension is controlled to produce substantially planar vertical and horizontal vitrified underground structures. Structures may be placed around a contaminated waste site to isolate the site or may be used as aquifer dikes.

Murphy, Mark T. (Kennewick, WA); Buelt, James L. (Richland, WA); Stottlemyre, James A. (Richland, WA); Tixier, Jr., John S. (Richland, WA)

1992-01-01T23:59:59.000Z

3

Outlooks of HLW Partitioning Technologies Usage for Recovering of Platinum Metals from Spent Fuel  

Science Conference Proceedings (OSTI)

The existing practice of management of high level waste (HLW) generated by NPPs, call for a task of selective separation of the most dangerous long-lived radionuclides with the purpose of their subsequent immobilization and disposal. HLW partitioning allows to reduce substantially the cost of vitrified product storage owing to isolation of the most dangerous radionuclides, such as transplutonium elements (TPE) into separate fractions of small volumes, intended for ultimate storage. By now numerous investigations on partitioning of HLW of various composition have been carried out in many countries and a lot of processes permitting to recover cesium, strontium, TPE and rare earth elements (REE) have been already tested. Apart from enumerated radionuclides, a fair quantity of palladium and rhodium presents in spent fuel, but the problem of these elements recovery has not yet been decided at the operating radiochemical plants. A negative effect of platinum group metals (PGM) occurrence is determined by the formation of separate metal phase, which not only worsens the conditions of glass-melting but also shortens considerably the service life of the equipment. At the same time, the exhaustion of PGMs natural resources may finally lead to such a growth of their costs that the spent nuclear fuel would became a substituting source of these elements industrial production. Allowing above mentioned, it is of interest to develop the technique for ''reactor'' palladium and rhodium recovery process which would be compatible with HLW partitioning and could be realized using the same facilities. In the report the data on platinum metals distribution in spent fuel reprocessing products and the several flowsheets for palladium separation from HLW are presented.

Pokhitonov, Y. A.; Estimantovskiy, V.; Romanovski, v.; Zatsev, B.; Todd, T.

2003-02-24T23:59:59.000Z

4

EM Waste Acceptance Product Specification (WAPS) for Vitrified High-Level Waste Forms  

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

EM Waste Acceptance Product EM Waste Acceptance Product Specification (WAPS) for Vitrified High-Level Waste Forms Presentation to the HLW Corporate Board July 24, 2008 By Tony Kluk/Ken Picha 2 Background * Originally Waste Acceptance Preliminary Specifications were Office of Civilian Radioactive Waste Management (RW) documents and project specific: - Defense Waste Processing Facility (PE-03, July 1989) - West Valley Demonstration Project (PE-04, January 1990) * Included many of same specifications as current version of WAPS * First version of RW Waste Acceptance System Requirements Document in January 1993 (included requirements for both SNF and HLW) * EM decided to extract requirements for HLW and put into the WAPS document 3 Background (Cont'd) * Lists technical specifications for acceptance of borosilicate HLW

5

Database and Interim Glass Property Models for Hanford HLW Glasses  

SciTech Connect

The purpose of this report is to provide a methodology for an increase in the efficiency and a decrease in the cost of vitrifying high-level waste (HLW) by optimizing HLW glass formulation. This methodology consists in collecting and generating a database of glass properties that determine HLW glass processability and acceptability and relating these properties to glass composition. The report explains how the property-composition models are developed, fitted to data, used for glass formulation optimization, and continuously updated in response to changes in HLW composition estimates and changes in glass processing technology. Further, the report reviews the glass property-composition literature data and presents their preliminary critical evaluation and screening. Finally the report provides interim property-composition models for melt viscosity, for liquidus temperature (with spinel and zircon primary crystalline phases), and for the product consistency test normalized releases of B, Na, and Li. Models were fitted to a subset of the screened database deemed most relevant for the current HLW composition region.

Hrma, Pavel R.; Piepel, Gregory F.; Vienna, John D.; Cooley, Scott K.; Kim, Dong-Sang; Russell, Renee L.

2001-07-24T23:59:59.000Z

6

HLW Glass Waste Loadings  

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

HLW HLW Glass Waste Loadings Ian L. Pegg Vitreous State Laboratory The Catholic University of America Washington, DC Overview Overview  Vitrification - general background  Joule heated ceramic melter (JHCM) technology  Factors affecting waste loadings  Waste loading requirements and projections  WTP DWPF  DWPF  Yucca Mountain License Application requirements on waste loading  Summary Vitrification  Immobilization of waste by conversion into a glass  Internationally accepted treatment for HLW  Why glass?  Amorphous material - able to incorporate a wide spectrum of elements over wide ranges of composition; resistant to radiation damage  Long-term durability - natural analogs Relatively simple process - amenable to nuclearization at large  Relatively simple process - amenable to nuclearization at large scale  There

7

PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)  

Science Conference Proceedings (OSTI)

The primary objective of this study is to demonstrate a mission scenario that uses pairwise and incidental blending of high level waste (HLW) to reduce the total mass of HLW glass. Secondary objectives include understanding how recent refinements to the tank waste inventory and solubility assumptions affect the mass of HLW glass and how logistical constraints may affect the efficacy of HLW blending.

CERTA, P.J.

2006-02-22T23:59:59.000Z

8

HLW Glass Studies: Development of Crystal-Tolerant HLW Glasses  

Science Conference Proceedings (OSTI)

In our study, a series of lab-scale crucible tests were performed on designed glasses of different compositions to further investigate and simulate the effect of Cr, Ni, Fe, Al, Li, and RuO2 on the accumulation rate of spinel crystals in the glass discharge riser of the HLW melter. The experimental data were used to expand the compositional region covered by an empirical model developed previously (Maty et al. 2010b), improving its predictive performance. We also investigated the mechanism for agglomeration of particles and impact of agglomerates on accumulation rate. In addition, the TL was measured as a function of temperature and composition.

Matyas, Josef; Huckleberry, Adam R.; Rodriguez, Carmen P.; Lang, Jesse B.; Owen, Antionette T.; Kruger, Albert A.

2012-04-02T23:59:59.000Z

9

Melter Testing with High Aluminum HLW Streams  

Hanford Tank Waste is High in Aluminum Estimated Al inventory is 8750 MT Problem: Large fraction of Al is in the HLW solids Greatly increases the ...

10

Vitrification and Product Testing of C-104 and AZ-102 Pretreated Sludge Mixed with Flowsheet Quantities of Secondary Wastes  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy (DOE) Office of River Protection (ORP) has acquired Hanford tank waste treatment services at a demonstration scale. The River Protection Project Waste Treatment Plant (RPP-WTP) team is responsible for producing an immobilized (vitrified) high-level waste (IHLW) waste form. Pacific Northwest National Laboratory, hereafter referred to as PNNL, has been contracted to produce and test a vitrified IHLW waste form from two Envelope D high-level waste (HLW) samples previously supplied to the RPP-WTP project by DOE.

Smith, Gary L.; Bates, Derrick J.; Goles, Ronald W.; Greenwood, Lawrence R.; Lettau, Ralph C.; Piepel, Gregory F.; Schweiger, Michael J.; Smith, Harry D.; Urie, Michael W.; Wagner, Jerome J.

2001-02-01T23:59:59.000Z

11

End of Year 2010 SNF & HLW Inventories  

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

Map of the United States of America that shows the location of approximately 64,000 MTHM of Spent Nuclear Fuel (SNF)& 275 High-Level Radioactive Waste (HLW) Canisters.

12

Petroleum Engineering Techniques for HLW Disposal  

Science Conference Proceedings (OSTI)

This paper describes why petroleum engineering techniques are of importance and can be used for underground disposal of HLW (high-level radioactive waste). It is focused on rock salt as a geological host medium in combination with disposal of the HLW canisters in boreholes drilled from the surface. Both permanent disposal and disposal with the option to retrieve the waste are considered. The paper starts with a description of the disposal procedure. Next disposal in deep boreholes is treated. Then the possible use of deviated boreholes and of multiple boreholes is discussed. Also waste isolation aspects and the implications of the HLW heat generation are treated. It appears that the use of deep boreholes can be beneficial, and also that--to a certain extent--borehole deviation offers possibilities. The benefits of using multiple boreholes are questionable for permanent disposal, while this technique cannot be applied for retrievable disposal. For the use of casing material, the additional temperature rise due to the HLW heat generation must be taken into account.

van den Broek, W. M. G. T.

2002-02-25T23:59:59.000Z

13

HLW Feed Delivery AZ101 Batch Transfer to the Private Contractor Transfer and Mixing Process Improvements [Initial Release at Rev 2  

SciTech Connect

The primary purpose of this business case is to provide Operations and Maintenance with a detailed transfer process review for the first High Level Waste (HLW) feed delivery to the Privatization Contractor (PC), AZ-101 batch transfer to PC. The Team was chartered to identify improvements that could be implemented in the field. A significant penalty can be invoked for not providing the quality, quantity, or timely delivery of HLW feed to the PC.

DUNCAN, G.P.

2000-02-28T23:59:59.000Z

14

HIGH ALUMINUM HLW GLASSES FOR HANFORDS WTP  

Science Conference Proceedings (OSTI)

The world's largest radioactive waste vitrification facility is now under construction at the United State Department of Energy's (DOE's) Hanford site. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is designed to treat nearly 53 million gallons of mixed hazardous and radioactive waste now residing in 177 underground storage tanks. This multi-decade processing campaign will be one of the most complex ever undertaken because of the wide chemical and physical variability of the waste compositions generated during the cold war era that are stored at Hanford. The DOE Office of River Protection (ORP) has initiated a program to improve the long-term operating efficiency of the WTP vitrification plants with the objective of reducing the overall cost of tank waste treatment and disposal and shortening the duration of plant operations. Due to the size, complexity and duration of the WTP mission, the lifecycle operating and waste disposal costs are substantial. As a result, gains in High Level Waste (HLW) and Low Activity Waste (LAW) waste loadings, as well as increases in glass production rate, which can reduce mission duration and glass volumes for disposal, can yield substantial overall cost savings. EnergySolutions and its long-term research partner, the Vitreous State Laboratory (VSL) of the Catholic University of America, have been involved in a multi-year ORP program directed at optimizing various aspects of the HLW and LAW vitrification flow sheets. A number of Hanford HLW streams contain high concentrations of aluminum, which is challenging with respect to both waste loading and processing rate. Therefore, a key focus area of the ORP vitrification process optimization program at EnergySolutions and VSL has been development of HLW glass compositions that can accommodate high Al{sub 2}O{sub 3} concentrations while maintaining high processing rates in the Joule Heated Ceramic Melters (JHCMs) used for waste vitrification at the WTP. This paper, reviews the achievements of this program with emphasis on the recent enhancements in Al{sub 2}O{sub 3} loadings in HLW glass and its processing characteristics. Glass formulation development included crucible-scale preparation and characterization of glass samples to assess compliance with all melt processing and product quality requirements, followed by small-scale screening tests to estimate processing rates. These results were used to down-select formulations for subsequent engineering-scale melter testing. Finally, further testing was performed on the DM1200 vitrification system installed at VSL, which is a one-third scale (1.20 m{sup 2}) pilot melter for the WTP HLW melters and which is fitted with a fully prototypical off-gas treatment system. These tests employed glass formulations with high waste loadings and Al{sub 2}O{sub 3} contents of {approx}25 wt%, which represents a near-doubling of the present WTP baseline maximum Al{sub 2}O{sub 3} loading. In addition, these formulations were processed successfully at glass production rates that exceeded the present requirements for WTP HLW vitrification by up to 88%. The higher aluminum loading in the HLW glass has an added benefit in that the aluminum leaching requirements in pretreatment are reduced, thus allowing less sodium addition in pretreatment, which in turn reduces the amount of LAW glass to be produced at the WTP. The impact of the results from this ORP program in reducing the overall cost and schedule for the Hanford waste treatment mission will be discussed.

KRUGER AA; JOSEPH I; BOWMAN BW; GAN H; KOT W; MATLACK KS; PEGG IL

2009-08-19T23:59:59.000Z

15

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

Science Conference Proceedings (OSTI)

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

Marra, J

2007-02-12T23:59:59.000Z

16

Thermal Diffusivity and Thermal Conductivity of HLW and LAW ...  

Science Conference Proceedings (OSTI)

In the present work, such data were collected for four waste glasses representative of those currently projected for treatment of Hanford HLW and LAW streams.

17

Advances in JHCM HLW Vitrification Technology through Scaled ...  

Science Conference Proceedings (OSTI)

... at Savannah River, WTP HLW and LAW at Hanford, as well Rokkasho in Japan . ... Waste at the Defense Waste Processing Facility through Sludge Batch 7b.

18

DOE Statement on Savannah River Site Vitrified Waste Concentrations |  

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

Statement on Savannah River Site Vitrified Waste Concentrations Statement on Savannah River Site Vitrified Waste Concentrations DOE Statement on Savannah River Site Vitrified Waste Concentrations April 30, 2010 - 12:30pm Addthis "The Office of Environmental Management has decided not to move forward at this time with its February decision to direct contractors to start planning for higher concentrations of plutonium in waste canisters at the Savannah River Site. While this may ultimately be a better way to manage and minimize the volume of waste, the Department wants to further review the issues involved before proceeding. No canisters have been filled at the higher concentration level." Addthis Related Articles Energy Secretary Chu Announces $6 Billion in Recovery Act Funding for Environmental Cleanup Department of Energy Projects Win 36 R&D 100 Awards for 2011

19

Summary - WTP HLW Waste Vitrification Facility  

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

W W HLW W DOE is Immob site's t facilitie Facility to iden the HL to be i norma The as along w Level ( * H * H * H Sy * Pu D The Ele Site: H roject: W Report Date: M ited States Waste T Why DOE Waste Vitrificatio s constructing bilization Plant tank wastes. T es including a H y (HLW). The ntify the critical LW and determ ncorporated in ally requires a T What th ssessment team with each elem (TRL) for the H LW Melter Fee LW Melter Pro LW Melter Offg ystem/Process ulse Jet Mixer isposal System To view the full T http://www.em.doe. objective of a Tech ements (CTEs), usin Hanford/ORP Waste Treatme March 2007 Departmen Treatmen W E-EM Did This n Facility a Waste Treat (WTP) at Hanf The WTP is com High-Level Wa purpose of this technology ele mine if these are to the final WT Technology Re he TRA Team m identified the

20

European experience in transport/storage cask for vitrified residues  

SciTech Connect

Available in abstract form only. Full text of publication follows: Because of the evolution of burnup of spent fuel to be reprocessed, the high activity vitrified residues would not be transported in the existing cask designs. Therefore, TN International has decided in the late nineties to develop a brand new design of casks with optimized capacity able to store and transport the most active and hottest canisters: the TN{sup TM}81 casks currently in use in Switzerland and the TN{sup TM}85 cask which shall permit in the near future in Germany the storage and the transport of the most active vitrified residues defining a thermal power of 56 kW (kilowatts). The challenges for the TN{sup TM}81 and TN{sup TM}85 cask designs were that the geometry entry data were very restrictive and were combined with a fairly wide range set by the AREVA NC Specification relative to vitrified residue canister. The TN{sup TM}81 and the TN{sup TM}85 casks have been designed to fully anticipate shipment constraints of the present vitrified residue production. It also used the feedback of current shipments and the operational constraints and experience of receiving and shipping facilities. The casks had to fit as much as possible in the existing procedures for the already existing flasks such as the TN{sup TM}28 cask and TS 28 V cask, all along the logistics chain of loading, unloading, transport and maintenance. (authors)

Otton, Camille; Sicard, Damien [AREVA - TN International (France)

2007-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

European experience in transport / storage cask for vitrified residues  

Science Conference Proceedings (OSTI)

Because of the evolution of burnup of spent fuel to be reprocessed, the high activity vitrified residues would not be transported in the existing cask designs. Therefore, TN International has decided in the late nineties to develop a brand new design of casks with optimized capacity able to store and transport the most active and hottest canisters: the TN{sup TM}81 casks currently in use in Switzerland and the TN{sup TM}85 cask which shall permit in the near future in Germany the storage and the transport of the most active vitrified residues defining a thermal power of 56 kW (kilowatts). The challenges for the TN{sup TM}81 and TN{sup TM}85 cask designs were that the geometry entry data were very restrictive and were combined with a fairly wide range set by the AREVA NC Specification relative to vitrified residue canister. The TN{sup TM}81 and the TN{sup TM}85 casks have been designed to fully anticipate shipment constraints of the present vitrified residue production. It also used the feedback of current shipments and the operational constraints and experience of receiving and shipping facilities. The casks had to fit as much as possible in the existing procedures for the already existing flasks such as the TN{sup TM}28 cask and TS 28 V cask, all along the logistics chain of loading, unloading, transport and maintenance. In addition, years of feedback and experience in design and operations - together with ever improved materials - have allowed finding further optimization of this type of cask design. In order to increase the loading capacity in terms of radioactive source terms and heat load by 40%, the cask design relies on innovative solutions and benchmarks from the current shipping campaigns. Currently, TN{sup TM}81 and TN{sup TM}85 are the only licensed casks that can transport and store 28 canisters with a total decay heat of 56 kW. It contributes to optimise the number of required transports to bring back high level waste residues to their producers. Three units have already been loaded and transported to ZWILAG (Zwischenlager Wuerenlingen AG) in Switzerland where they are stored for 40 years. Based on the same design but integrating the German Authorities and German users specificities, the TN{sup TM}85 cask is dedicated to the transport and storage of vitrified residues to Germany. It is presently at the final licensing stage. The transport cask approval expertise has now been granted, and the storage expertise is in the final steps. The first transport with TN{sup TM}85 cask is scheduled up to now in 2007 and the commissioning operations are under preparation. These two casks are key elements for the whole reprocessing system of AREVA as they enable the transport and the storage of the vitrified residues. (authors)

Blachet, L.; Otton, C.; Sicard, D. [AREVA TN International (France)

2007-07-01T23:59:59.000Z

22

Idaho Waste Vitrification Facilities Project Vitrified Waste Interim Storage Facility  

SciTech Connect

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

Bonnema, Bruce Edward

2001-09-01T23:59:59.000Z

23

Water Quantity | Open Energy Information  

Open Energy Info (EERE)

Quantity Jump to: navigation, search Retrieved from "http:en.openei.orgwindex.php?titleWaterQuantity&oldid612364...

24

DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES  

Science Conference Proceedings (OSTI)

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 borosilicate glass or phosphate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt waste plus glass forming frit additives and cast. A second reason that glass has become widely used for HLW is that the short range order (SRO) and medium range order (MRO) found in glass atomistically bonds the radionuclides and governs the melt properties such as viscosity, resistivity, sulphate solubility. The molecular structure of glass controls contaminant/radionuclide release by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to waste variability. Nuclear waste glasses melt between 1050-1150 C which minimizes the volatility of radioactive components such as Tc{sup 99}, Cs{sup 137}, and I{sup 129}. Nuclear waste glasses have good long term stability including irradiation resistance. Process control models based on the molecular structure of glass have been mechanistically derived and have been demonstrated to be accurate enough to control the world's largest HLW Joule heated ceramic melter in the US since 1996 at 95% confidence.

Jantzen, C.

2010-03-18T23:59:59.000Z

25

Standard test methods for vitrified ceramic materials for electrical applications  

E-Print Network (OSTI)

1.1 These test methods outline procedures for testing samples of vitrified ceramic materials that are to be used as electrical insulation. Where specified limits are mentioned herein, they shall not be interpreted as specification limits for completed insulators. 1.2 These test methods are intended to apply to unglazed specimens, but they may be equally suited for testing glazed specimens. The report section shall indicate whether glazed or unglazed specimens were tested. 1.3 The test methods appear as follows: This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. Specific precaution statements are given in 11.3, 13.5, and 15.3.

American Society for Testing and Materials. Philadelphia

1986-01-01T23:59:59.000Z

26

Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility  

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

6 6 Technology Readiness Assessment for the Waste Treatment and Immobilization Plant (WTP) HLW Waste Vitrification Facility L. Holton D. Alexander C. Babel H. Sutter J. Young August 2007 Prepared by the U.S. Department of Energy Office of River Protection Richland, Washington, 99352 07-DESIGN-046 Technology Readiness Assessment for the Waste Treatment and Immobilization Plant (WTP) HLW Waste Vitrification Facility L. Holton D. Alexander C. Babel H. Sutter J. Young August 2007 Prepared by the U.S. Department of Energy Office of River Protection under Contract DE-AC05-76RL01830 07-DESIGN-046 iii Summary The U.S. Department of Energy (DOE), Office of River Protection (ORP) and the DOE Office of Environmental and Radioactive Waste Management (EM), Office of Project Recovery have completed a

27

Quantity | Open Energy Information  

Open Energy Info (EERE)

View View New Pages Recent Changes All Special Pages Semantic Search/Querying Get Involved Help Apps Datasets Community Login | Sign Up Search Special page Facebook icon Twitter icon » Quantity Jump to: navigation, search Properties of type "Quantity" Showing 53 properties using this type. A Property:Area Property:AvgReservoirDepth C Property:Capacity E Property:EstReservoirVol Property:EstimatedCostHighUSD Property:EstimatedCostLowUSD Property:EstimatedCostMedianUSD Property:EstimatedTime Property:EstimatedTimeHigh Property:EstimatedTimeLow Property:EstimatedTimeMedian F Property:FirstWellDepth Property:FirstWellFlowRate G Property:GeneratingCapacity Property:GrossProdCapacity I Property:IdentifiedHydrothermalPotential Property:InstalledCapacity M Property:MeanCapacity N Property:NetProdCapacity

28

Long-term management of high-level radioactive waste (HLW) and...  

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

HLW is the highly radioactive material resulting from the reprocessing of SNF. Under the Nuclear Waste Policy Act of 1982, the federal government is responsible for the disposal...

29

Melter Throughput Enhancements for High-Iron HLW  

Science Conference Proceedings (OSTI)

This report describes work performed to develop and test new glass and feed formulations in order to increase glass melting rates in high waste loading glass formulations for HLW with high concentrations of iron. Testing was designed to identify glass and melter feed formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts to assess melt rate using a vertical gradient furnace system and to develop new formulations with enhanced melt rate. Testing evaluated the effects of waste loading on glass properties and the maximum waste loading that can be achieved. The results from crucible-scale testing supported subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass and feed formulations on waste processing rate and product quality. The DM100 was selected as the platform for these tests due to its extensive previous use in processing rate determination for various HLW streams and glass compositions.

Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Gan, Hoa [The Catholic University of America, Washington, DC (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Chaudhuri, Malabika [The Catholic University of America, Washington, DC (United States); Kot, Wing [The Catholic University of America, Washington, DC (United States)

2012-12-26T23:59:59.000Z

30

EMPIRICAL MODEL FOR FORMULATION OF CRYSTAL-TOLERANT HLW GLASSES  

Science Conference Proceedings (OSTI)

Historically, high-level waste (HLW) glasses have been formulated with a low liquideus temperature (T{sub L}), or temperature at which the equilibrium fraction of spinel crystals in the melt is below 1 vol % (T{sub 0.01}), nominally below 1050 C. These constraints cannot prevent the accumulation of large spinel crystals in considerably cooler regions ({approx} 850 C) of the glass discharge riser during melter idling and significantly limit the waste loading, which is reflected in a high volume of waste glass, and would result in high capital, production, and disposal costs. A developed empirical model predicts crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass, and thereby provides guidance in formulating crystal-tolerant glasses that would allow high waste loadings by keeping the spinel crystals small and therefore suspended in the glass.

KRUGER AA; MATYAS J; HUCKLEBERRY AR; VIENNA JD; RODRIGUEZ CA

2012-03-07T23:59:59.000Z

31

Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections  

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

Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel Task: Identify Shortline Railroads Serving Nuclear Power Plants Establish Contact Information with Railroads Officials Field Review of each Railroad's Physical and Operational Infrastructure Facilitate Upgrades to Meet Safe Acceptable Standards Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel More Documents & Publications TEC Meeting Summaries - February 2008 Presentations TEC Meeting Summaries - July 2007 Presentations TEC Meeting Summaries - September 2006

32

Redox Control For Hanford HLW Feeds VSL-12R2530-1, REV 0  

SciTech Connect

The principal objectives of this work were to investigate the effects of processing simulated Hanford HLW at the estimated maximum concentrations of nitrates and oxalates and to identify strategies to mitigate any processing issues resulting from high concentrations of nitrates and oxalates. This report provides results for a series of tests that were performed on the DM10 melter system with simulated C-106/AY-102 HLW. The tests employed simulated HLW feeds containing variable amounts of nitrates and waste organic compounds corresponding to maximum concentrations proj ected for Hanford HLW streams in order to determine their effects on glass production rate, processing characteristics, glass redox conditions, melt pool foaming, and the tendency to form secondary phases. Such melter tests provide information on key process factors such as feed processing behavior, dynamic effects during processing, processing rates, off-gas amounts and compositions, foaming control, etc., that cannot be reliably obtained from crucible melts.

Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States)

2012-12-13T23:59:59.000Z

33

Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF)  

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

GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor...

34

Status of Sandia HLW canister/overpack program studies  

DOE Green Energy (OSTI)

The focus of the Sandia program has been to identify an extended life alloy suitable as an overpack surrounding an HLW canister. The function of the overpack, which may be only millimeters thick, is corrosion resistance, not support strength. Laboratory and hot-cell tests are being used to measure the corrosion rates and assess the metallurgical behavior of selected engineered barrier materials. Field and in situ tests and demonstrations are in the planning stage. Recent experimental results are reviewed, and the status of the various phases of this program are described. Several candidate alloys have been examined for corrosion behavior under environmental conditions typical of a salt repository. The prime candidate for long-lived overpacks, TiCode-12, has not been disqualified by any of the tests and overtests conducted in our investigations. However further testing of potential failure mechanisms are being evaluated before final material selection is made. Nickel-based and lower-cost alloys will also be examined. This program will culminate with large-scale overpack fabrication demonstrations and field testing in salt.

Molecke, M.A.; Abrego, L.

1980-01-01T23:59:59.000Z

35

Safety analysis report vitrified high level waste type B shipping cask  

Science Conference Proceedings (OSTI)

This Safety Analysis Report describes the design, analyses, and principle features of the Vitrified High Level Waste (VHLW) Cask. In preparing this report a detailed evaluation of the design has been performed to ensure that all safety, licensing, and operational goals for the cask and its associated Department of Energy program can be met. The functions of this report are: (1) to fully document that all functional and regulatory requirements of 10CFR71 can be met by the package; and (2) to document the design and analyses of the cask for review by the Nuclear Regulatory Commission. The VHLW Cask is the reusable shipping package designed by GNSI under Department of Energy contract DE-AC04-89AL53-689 for transportation of Vitrified High Level Waste, and to meet the requirements for certification under 10CFR71 for a Type B(U) package. The VHLW cask has been designed as packaging for transport of canisters of Vitrified High Level Waste solidified at Department of Energy facilities.

NONE

1995-03-01T23:59:59.000Z

36

A Conjoint Model of Quantity Discounts  

Science Conference Proceedings (OSTI)

Quantity discount pricing is a common practice used by business-to-business and business-to-consumer companies. A key characteristic of quantity discount pricing is that the marginal price declines with higher purchase quantities. In this paper, we propose ... Keywords: choice models, conjoint analysis, mixed logit, quantity discounts, willingness to pay

Raghuram Iyengar; Kamel Jedidi

2012-03-01T23:59:59.000Z

37

FINAL REPORT TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-02R0100-2 REV 1 2/17/03  

Science Conference Proceedings (OSTI)

This document provides the final report on data and results obtained from a series of nine tests performed on the one-third scale DuraMelter{trademark} 1200 (DM1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part B1 [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plans. The nine tests reported here were preceded by an initial series of short-duration tests conducted to support the start-up and commissioning of this system. This report is a followup to the previously issued Preliminary Data Summary Reports. The DM1200 system was deployed for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. These tests include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The primary objective of the present series of tests was to determine the effects of a variety of parameters on the glass production rate in comparison to the RPP-WTP HL W design basis of 400 kg/m{sup 2}/d. Previous testing on the DMIOOO system [1] concluded that achievement of that rate with simulants of projected WTP melter feeds (AZ-101 and C-106/AY-102) was unlikely without the use of bubblers. As part of those tests, the same feed that was used during the cold-commissioning of the West Valley Demonstration Project (WVDP) HLW vitrification system was run on the DM1000 system. The DM1000 tests reproduced the rates that were obtained at the larger WVDP facility, lending confidence to the tests results [1]. Since the inclusion or exclusion of a bubbler has significant design implications, the Project commissioned further tests to address this issue. In an effort to identify factors that might increase the glass production rate for projected WTP melter feeds, a subsequent series of tests was performed on the DM100 system. Several tests variables led to glass production rate increases to values significantly above the 400 kg/m2/d requirement. However, while small-scale melter tests are useful for screening relative effects, they tend to overestimate absolute glass production rates, particularly for un-bubbled tests. Consequently, when scale-up effects were taken into account, it was not clear that any of the variables investigated would conclusively meet the 400 kg/m{sup 2}/d requirement without bubbling. The present series of tests was therefore performed on the DM1200 one-third scale HLW pilot melter system to provide the required basis for a final decision on whether bubblers would be included in the HLW melter. The present tests employed the same AZ-101 waste simulant and glass composition that was used for previous testing for consistency and comparability with the results from the earlier tests.

KRUGER AA; MATLACK KS; KOT WK; BARDAKCI T; GONG W; D'ANGELO NA; SCHATZ TR; PEGG IL

2011-12-29T23:59:59.000Z

38

Disposal of defense spent fuel and HLW from the Idaho Chemical Processing Plant  

SciTech Connect

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

Ermold, L.F.; Loo, H.H.; Klingler, R.D.; Herzog, J.D.; Knecht, D.A.

1992-12-01T23:59:59.000Z

39

HLW Salt Disposition Alternatives Identification Preconceptual Phase I Summary Report (Including Attachments)  

SciTech Connect

The purpose of this report is to summarize the process used by the Team to systematically develop alternative methods or technologies for final disposition of HLW salt. Additionally, this report summarizes the process utilized to reduce the total list of identified alternatives to an ''initial list'' for further evaluation. This report constitutes completion of the team charter major milestone Phase I Deliverable.

Piccolo, S.F.

1999-07-09T23:59:59.000Z

40

Title: An Advanced Solution for the Storage, Transportation and Disposal of Vitrified High Level Waste  

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

Presented at Global 99, Jackson, Wyoming, August 29 - September 2, 1999 Presented at Global 99, Jackson, Wyoming, August 29 - September 2, 1999 1 AN ADVANCED SOLUTION FOR THE STORAGE, TRANSPORTATION AND DISPOSAL OF SPENT FUEL AND VITRIFIED HIGH LEVEL WASTE William J. Quapp Teton Technologies, Inc. 860 W. Riverview Dr. Idaho Falls, ID 83401 208-535-9001 ABSTRACT For future nuclear power deployment in the US, certain changes in the back end of the fuel cycle, i.e., disposal of high level waste and spent fuel, must become a real options. However, there exists another problem from the front end of the fuel cycle which has until recently, received less attention. Depleted uranium hexafluoride is a by-product of the enrichment process and has accumulated for over 50 years. It now represents a potential environmental problem. This paper describes a

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41

Microsoft PowerPoint - 2-05 PEGG-2 - Melter Tests with High Al HLW - Nov 2010 emb.ppt  

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

Melter Melter Testing with High Aluminum HLW Streams Ian L. Pegg, Hao Gan, Wing K. Kot, Keith S. Matlack, and Innocent Joseph * Vitreous State Laboratory The Catholic University of America Washington, DC * EnergySolutions, Inc. DOE EM Waste Processing Technical Exchange 2010 Print Close Melter Testing with High Aluminum HLW Streams 2 LAW Vitrification (90+% of waste mass) HLW Vitrification (90+% of waste activity) Pretreatment (solid/liquid separation, Cs-IX, Al, Cr, leaching) SLUDGE SUPERNATE Maximize Mass Maximize Activity Hanford WTP - Key Process Flows LAW glass disposed on site HLW glass disposed of in National Geologic Repository - TBD * Supernate: Solution of Na, Al, P, K, S, Cl, Cs, Tc, nitrates, hydroxides... * Sludge: Solids high in Fe, Al, Zr, Cr, Bi, Sr, TRU, oxides, hydroxides....

42

SEISMIC DESIGN EVALUATION GUIDELINES FOR BURIED PIPING FOR THE DOE HLW FACILITIES'  

Office of Scientific and Technical Information (OSTI)

6 1 6 1 7 1 1 SEISMIC DESIGN EVALUATION GUIDELINES FOR BURIED PIPING FOR THE DOE HLW FACILITIES' Chi-Wen Lin Consultant, Martinez, CA George Antaki Westinghouse Savannah River Co., Aiken, SC Kamal Bandyopadhyay Brookhaven National Lab., Upton, NY ABSTRACT This paper presents the seismic design and evaluation guidelines for underground piping for the Department of Energy (DOE) High-Level-Waste (HLW) Facilities. The underground piping includes both single and double containment steel pipes and concrete pipes with steel lining, with particular emphasis on the double containment piping. The design and evaluation guidelines presented in this paper follow the generally accepted beam-on-elastic- foundation analysis principle and the inertial response calculation method, respectively, for piping directly

43

Shielding analysis of the TRUPACT-series casks for transportation of Hanford HLW  

SciTech Connect

In this paper, the authors propose the possibility of utilizing the TRUPACT-series casks for the transportation of high-level waste (HLW) from the Hanford reservation. The configurations of the TRUPACT series are a rectangular parallelepiped and a right circular cylinder, which are the TRUPACT-1 and -11, respectively. The TRUPACT series was designed as a type B contact-handled transuranic (CH-TRU) waste transportation system for use in Waste Isolation Pilot Plant-related operations and was subjected to type B container accident tests, which it successfully passed. Thus from a safety standpoint, the TRUPACT series is provided with double containment, impact limitation, and fire-retardant capabilities. However, the shielding analysis has shown the major modifications are required to allow for the transport of even a reasonable fraction of Hanford HLW.

Banjac, V.; Sanchez, P.E.; Hills, C.R.; Heger, A.S. (Univ. of New Mexico, Albuquerque, NM (United States))

1993-01-01T23:59:59.000Z

44

HLW MELTER CONTROL STRATEGY WITHOUT VISUAL FEEDBACK VSL-12R2500-1 REV 0  

Science Conference Proceedings (OSTI)

Plans for the treatment of high level waste (HL W) at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) are based upon the inventory of the tank wastes, the anticipated performance of the pretreatment processes, and current understanding of the capability of the borosilicate glass waste form [I]. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat and mass transfer and increase glass melting rates. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of ~ 1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HL W waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage.

KRUGER AA; JOSPEH I; MATLACK KS; CALLOW RA; ABRAMOWITZ H; PEGG IL; BRANDYS M; KOT WK

2012-11-13T23:59:59.000Z

45

HLW Melter Control Strategy Without Visual Feedback VSL-12R2500-1 Rev 0  

SciTech Connect

Plans for the treatment of high level waste (HL W) at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) are based upon the inventory of the tank wastes, the anticipated performance of the pretreatment processes, and current understanding of the capability of the borosilicate glass waste form [I]. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat and mass transfer and increase glass melting rates. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of ~ 1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HL W waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150?C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage.

Kruger, A A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Callow, Richard A. [The Catholic University of America, Washington, DC (United States); Abramowitz, Howard [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Brandys, Marek [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States)

2012-11-13T23:59:59.000Z

46

HIGH ALUMINUM HLW (HIGH LEVEL WASTE ) GLASSES FOR HANFORDS WTP (WASTE TREATMENT PROJECT)  

Science Conference Proceedings (OSTI)

This paper presents the results of glass formulation development and melter testing to identify high waste loading glasses to treat high-Al high level waste (HLW) at Hanford. Previous glass formulations developed for this HLW had high waste loadings but their processing rates were lower that desired. The present work was aimed at improving the glass processing rate while maintaining high waste loadings. Glass formulations were designed, prepared at crucible-scale and characterized to determine their properties relevant to processing and product quality. Glass formulations that met these requirements were screened for melt rates using small-scale tests. The small-scale melt rate screening included vertical gradient furnace (VGF) and direct feed consumption (DFC) melter tests. Based on the results of these tests, modified glass formulations were developed and selected for larger scale melter tests to determine their processing rate. Melter tests were conducted on the DuraMelter 100 (DMIOO) with a melt surface area of 0.11 m{sup 2} and the DuraMelter 1200 (DMI200) HLW Pilot Melter with a melt surface area of 1.2 m{sup 2}. The newly developed glass formulations had waste loadings as high as 50 wt%, with corresponding Al{sub 2}O{sub 3} concentration in the glass of 26.63 wt%. The new glass formulations showed glass production rates as high as 1900 kg/(m{sup 2}.day) under nominal melter operating conditions. The demonstrated glass production rates are much higher than the current requirement of 800 kg/(m{sup 2}.day) and anticipated future enhanced Hanford Tank Waste Treatment and Immobilization Plant (WTP) requirement of 1000 kg/(m{sup 2}.day).

KRUGER AA; BOWAN BW; JOSEPH I; GAN H; KOT WK; MATLACK KS; PEGG IL

2010-01-04T23:59:59.000Z

47

Collaboration, Automation, and Information Management at Hanford High Level Radioactive Waste (HLW) Tank Farms - 14210  

Science Conference Proceedings (OSTI)

Washington River Protection Solutions (WRPS), operator of High Level Radioactive Waste (HLW) Tank Farms at the Hanford Site, is taking an over 20-year leap in technology, replacing systems that were monitored with clipboards and obsolete computer systems, as well as solving major operations and maintenance hurdles in the area of process automation and information management. While WRPS is fully compliant with procedures and regulations, the current systems are not integrated and do not share data efficiently, hampering how information is obtained and managed.

Aurah, Mirwaise Y.; Roberts, Mark A.

2013-12-12T23:59:59.000Z

48

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

Science Conference Proceedings (OSTI)

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

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

2010-01-04T23:59:59.000Z

49

Method and Apparatus for Measuring Radiation Quantities  

DOE Patents (OSTI)

This patent application describes a compact dosimeter for measuring X-ray and gamma radiation by the use of solutions which undergo a visible color change upon exposure to a predetermined quantity of radiation.

Roberts, N.O.

1950-07-31T23:59:59.000Z

50

The Production of Advanced Glass Ceramic HLW Forms using Cold Crucible Induction Melter  

SciTech Connect

Cold Crucible Induction Melters (CCIMs) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in the 21st century. Unlike the existing Joule-Heated Melters (JHMs) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIMs offer unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. This paper discusses advantageous features of the CCIM, with emphasis on features that overcome the historical issues with the JHMs presently utilized, as well as the benefits of glass ceramic waste forms over borosilicate glass waste forms. These advantages are then validated based on recent INL testing to demonstrate a first-of-a-kind formulation of a non-radioactive ceramic-based waste form utilizing a CCIM.

Veronica J Rutledge; Vince Maio

2013-10-01T23:59:59.000Z

51

Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel  

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

Transportation Transportation Stakeholders National Transportation Stakeholders National Transportation Stakeholders National Transportation Stakeholders Forum Forum 2011 Annual Meeting 2011 Annual Meeting 2011 Annual Meeting 2011 Annual Meeting May 11, 2011 May 11, 2011 Evaluation of Shortline Railroads Evaluation of Shortline Railroads & & & & SNF/HLW Rail Shipment Inspections SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel Tasked for the Transportation of Spent Nuclear Fuel Evaluation of Shortline Railroads Evaluation of Shortline Railroads Evaluation of Shortline Railroads Evaluation of Shortline Railroads Task: Task: Task: Task: Identify Shortline Railroads Serving Nuclear Power Plants Identify Shortline Railroads Serving Nuclear Power Plants

52

Property:DayQuantity | Open Energy Information  

Open Energy Info (EERE)

DayQuantity DayQuantity Jump to: navigation, search Property Name DayQuantity Property Type String Description Enter the number of days (the default), but convert it to whatever time metric you'd like. Please note that the conversion to months and years is not accurate since the conversion depends on the specific years and months, but which are not known. Acceptable units (and their conversions) are: 1 day,Day,days,Days,DAY,DAYS,d,D 24 hour,hours,Hour,Hours,hr,hrs,HOUR,HOURS,HR,HRS 1440 minute,minutes,Minute,Minutes,min,Min,MINUTE,MINUTES,MIN 86400 second,seconds,Second,Seconds,sec,Sec,SECOND,SECONDS,SEC 0.142857143 week,weeks,Week,Weeks,wk,Wk,WEEK,WEEKS,WK 0.032786885 month,months,Month,Months,MONTH,MONTHS 0.002739726 year,years,Year,Years,yr,Yr,YEAR,YEARS,YR 1 day,Day,days,Days,DAY,DAYS,d,D

53

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

SciTech Connect

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

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

2009-12-30T23:59:59.000Z

54

INTEGRATED DM 1200 MELTER TESTING OF HLW C-106/AY-102 COMPOSITION USING BUBBLERS VSL-03R3800-1 REV 0 9/15/03  

SciTech Connect

This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of simulated HLW C-106/AY-102 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW C-106/AY-102 feed; determine the effect of bubbling rate on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post test inspections of system components.

KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D'ANGELO NA; KOT WK; PEGG IL

2011-12-29T23:59:59.000Z

55

Utah Quantity of Production Associated with Reported Wellhead...  

Annual Energy Outlook 2012 (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Utah Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade...

56

Missouri Quantity of Production Associated with Reported Wellhead...  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Missouri Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet)...

57

Tennessee Quantity of Production Associated with Reported Wellhead...  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Tennessee Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet)...

58

Illinois Quantity of Production Associated with Reported Wellhead...  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Illinois Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet)...

59

Arizona Quantity of Production Associated with Reported Wellhead...  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Arizona Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet)...

60

Florida Quantity of Production Associated with Reported Wellhead...  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Florida Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet)...

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

Kansas Quantity of Production Associated with Reported Wellhead...  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Kansas Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet)...

62

U.S. Quantity of Production Associated with Reported Wellhead...  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) U.S. Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade...

63

Maryland Quantity of Production Associated with Reported Wellhead...  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Maryland Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet)...

64

FINAL REPORT START-UP AND COMMISSIONING TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-01R0100-2 REV 0 1/20/03  

Science Conference Proceedings (OSTI)

This document provides the final report on data and results obtained from commissioning tests performed on the one-third scale DuraMelter{trademark} 1200 (DM 1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part BI [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plan. This report is a followup to the previously issued Preliminary Data Summary Report. The DM1200 system will be used for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. This will include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The results presented in this report are from the initial series of short-duration tests that were conducted to support the start-up and commissioning of this system prior to conducting the main body of development tests that have been planned for this system. These tests were directed primarily at system 'debugging,' operator training, and procedure refinement. The AZ-101 waste simulant and glass composition that was used for previous testing was selected for these tests.

KRUGER AA; MATLACK KS; KOT WK; BRANDYS M; WILSON CN; SCHATZ TR; GONG W; PEGG IL

2011-12-29T23:59:59.000Z

65

Final Report - Sulfate Solubility in RPP-WTP HLW Glasses, VSL-06R6780-1, Rev. 0  

SciTech Connect

This report describes the results of work and testing specified by Test Specifications 24590-HLW-TSP-RT-01-006 Rev 1, Test Plans VSL-02T7800-1 Rev 1 and Test Exceptions 24590-HLW-TEF-RT-05-00007. The work and any associated testing followed established quality assurance requirements and were conducted as authorized. The descriptions provided in this report are an accurate account of both the conduct of the work and the data collected. Results required by the Test Plans are reported. Also reported are any unusual or anomalous occurrences that are different from the starting hypotheses. The test results and this report have been reviewed and verified.

Kruger, Albert A.; Pegg, I. L.; Feng, A.; Gan, H.; Kot, W. K.

2013-12-03T23:59:59.000Z

66

Successful Use of Remote Engineering Technology to Upgrade Electrical Power Supplies to a Plant Producing Vitrified Highly Active Waste  

Science Conference Proceedings (OSTI)

This paper describes a remote handling intervention project on the Sellafield site in the UK that successfully replaced a critical part of a critical plant in a highly radioactive and contaminated cell. The aim of the project was to replace the existing design of electrical power supplies inside the plant that vitrifies high level liquid waste with a new improved design. The project designed and built a hydraulic manipulator and associated work-heads and tooling to be deployed in cell to remotely replace the power supplies. As part of this replacement process, the project also designed and built a drilling rig to remotely drill holes through the cell wall suitable for the new design of electrical power supplies. (authors)

Harken, J.P. [Nexia Solutions Ltd, Workington, Cumbria CA (United Kingdom)

2007-07-01T23:59:59.000Z

67

Enhanced sludge processing of HLW: Hydrothermal oxidation of chromium, technetium, and complexants by nitrate. 1997 mid-year progress report  

SciTech Connect

'Treatment of High Level Waste (HLW) is the second most costly problem identified by OEM. In order to minimize costs of disposal, the volume of HLW requiring vitrification and long term storage must be reduced. Methods for efficient separation of chromium from waste sludges, such as the Hanford Tank Wastes (HTW), are key to achieving this goal since the allowed level of chromium in high level glass controls waste loading. At concentrations above 0.5 to 1.0 wt.% chromium prevents proper vitrification of the waste. Chromium in sludges most likely exists as extremely insoluble oxides and minerals, with chromium in the plus III oxidation state [1]. In order to solubilize and separate it from other sludge components, Cr(III) must be oxidized to the more soluble Cr(VI) state. Efficient separation of chromium from HLW could produce an estimated savings of $3.4B[2]. Additionally, the efficient separation of technetium [3], TRU, and other metals may require the reformulation of solids to free trapped species as well as the destruction of organic complexants. New chemical processes are needed to separate chromium and other metals from tank wastes. Ideally they should not utilize additional reagents which would increase waste volume or require subsequent removal. The goal of this project is to apply hydrothermal processing for enhanced chromium separation from HLW sludges. Initially, the authors seek to develop a fundamental understanding of chromium speciation, oxidation/reduction and dissolution kinetics, reaction mechanisms, and transport properties under hydrothermal conditions in both simple and complex salt solutions. The authors also wish to evaluate the potential of hydrothermal processing for enhanced separations of technetium and TRU by examining technetium and TRU speciation at hydrothermal conditions optimal for chromium dissolution.'

Buelow, S.

1997-06-01T23:59:59.000Z

68

Final Report - Crystal Settling, Redox, and High Temperature Properties of ORP HLW and LAW Glasses, VSL-09R1510-1, Rev. 0, dated 6/18/09  

Science Conference Proceedings (OSTI)

The radioactive tank waste treatment programs at the U. S. Department of Energy (DOE) have featured joule heated ceramic melter technology for the vitrification of high level waste (HLW). The Hanford Tank Waste Treatment and Immobilization Plant (WTP) employs this same basic technology not only for the vitrification of HLW streams but also for the vitrification of Low Activity Waste (LAW) streams. Because of the much greater throughput rates required of the WTP as compared to the vitrification facilities at the West Valley Demonstration Project (WVDP) or the Defense Waste Processing Facility (DWPF), the WTP employs advanced joule heated melters with forced mixing of the glass pool (bubblers) to improve heat and mass transport and increase melting rates. However, for both HLW and LAW treatment, the ability to increase waste loadings offers the potential to significantly reduce the amount of glass that must be produced and disposed and, therefore, the overall project costs. This report presents the results from a study to investigate several glass property issues related to WTP HLW and LAW vitrification: crystal formation and settling in selected HLW glasses; redox behavior of vanadium and chromium in selected LAW glasses; and key high temperature thermal properties of representative HLW and LAW glasses. The work was conducted according to Test Plans that were prepared for the HLW and LAW scope, respectively. One part of this work thus addresses some of the possible detrimental effects due to considerably higher crystal content in waste glass melts and, in particular, the impact of high crystal contents on the flow property of the glass melt and the settling rate of representative crystalline phases in an environment similar to that of an idling glass melter. Characterization of vanadium redox shifts in representative WTP LAW glasses is the second focal point of this work. The third part of this work focused on key high temperature thermal properties of representative WTP HLW and LAW glasses over a wide range of temperatures, from the melter operating temperature to the glass transition.

Kruger, Albert A.; Wang, C.; Gan, H.; Pegg, I. L.; Chaudhuri, M.; Kot, W.; Feng, Z.; Viragh, C.; McKeown, D. A.; Joseph, I.; Muller, I. S.; Cecil, R.; Zhao, W.

2013-11-13T23:59:59.000Z

69

Final Report - Testing of Optimized Bubbler Configuration for HLW Melter VSL-13R2950-1, Rev. 0, dated 6/12/2013  

Science Conference Proceedings (OSTI)

The principal objective of this work was to determine the glass production rate increase and ancillary effects of adding more bubbler outlets to the current WTP HLW melter baseline. This was accomplished through testing on the HLW Pilot Melter (DM1200) at VSL. The DM1200 unit was selected for these tests since it was used previously with several HLW waste streams including the four tank wastes proposed for initial processing at Hanford. This melter system was also used for the development and optimization of the present baseline WTP HLW bubbler configuration for the WTP HLW melter, as well as for MACT testing for both HLW and LAW. Specific objectives of these tests were to: Conduct DM1200 melter testing with the baseline WTP bubbling configuration and as augmented with additional bubblers. Conduct DM1200 melter testing to differentiate the effects of total bubbler air flow and bubbler distribution on glass production rate and cold cap formation. Collect melter operating data including processing rate, temperatures at a variety of locations within the melter plenum space, melt pool temperature, glass melt density, and melter pressure with the baseline WTP bubbling configuration and as augmented with additional bubblers. Collect melter exhaust samples to compare particulate carryover for different bubbler configurations. Analyze all collected data to determine the effects of adding more bubblers to the WTP HLW melter to inform decisions regarding future lid re-designs. The work used a high aluminum HLW stream composition defined by ORP, for which an appropriate simulant and high waste loading glass formulation were developed and have been previously processed on the DM1200.

Kruger, Albert A.; Pegg, I. L.; Callow, R. A.; Joseph, I.; Matlack, K. S.; Kot, W. K.

2013-11-13T23:59:59.000Z

70

Final Report - Management of High Sulfur HLW, VSL-13R2920-1, Rev. 0, dated 10/31/2013  

Science Conference Proceedings (OSTI)

The present report describes results from a series of small-scale crucible tests to determine the extent of corrosion associated with sulfur containing HLW glasses and to develop a glass composition for a sulfur-rich HLW waste stream, which was then subjected to small-scale melter testing to determine the maximum acceptable sulfate loadings. In the present work, a new glass formulation was developed and tested for a projected Hanford HLW composition with sulfate concentrations high enough to limit waste loading. Testing was then performed on the DM10 melter system at successively higher waste loadings to determine the maximum waste loading without the formation of a separate sulfate salt phase. Small scale corrosion testing was also conducted using the glass developed in the present work, the glass developed in the initial phase of this work [26], and a high iron composition, all at maximum sulfur concentrations determined from melter testing, in order to assess the extent of Inconel 690 and MA758 corrosion at elevated sulfate contents.

Kruger, Albert A.; Gan, H.; Pegg, I. L.; Feng, Z.; Gan, H,; Joseph, I.; Matlack, K. S.

2013-11-13T23:59:59.000Z

71

IMPACT OF PARTICLE AGGLOMERATION ON ACCUMULATION RATES IN THE GLASS DISCHARGE RISER OF HLW MELTER  

SciTech Connect

The major factor limiting waste loading in continuous high-level radioactive waste (HLW) melters is an accumulation of particles in the glass discharge riser during a frequent and periodic idling of more than 20 days. An excessive accumulation can produce robust layers a few centimeters thick, which may clog the riser, preventing molten glass from being poured into canisters. Since the accumulation rate is driven by the size of particles we investigated with x-ray microtomography, scanning electron microscopy, and image analysis the impact of spinel forming components, noble metals, and alumina on the size, concentration, and spatial distribution of particles, and on the accumulation rate. Increased concentrations of Fe and Ni in the baseline glass resulted in the formation of large agglomerates that grew over the time to an average size of ~185155 {micro}m, and produced >3 mm thick layer after 120 h at 850C. The noble metals decreased the particle size, and therefore significantly slowed down the accumulation rate. Addition of alumina resulted in the formation of a network of spinel dendrites which prevented accumulation of particles into compact layers.

Kruger AA; Rodriguez CA: Matyas J; Owen AT; Jansik DP; Lang JB

2012-11-12T23:59:59.000Z

72

Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities  

SciTech Connect

This paper presents the seismic design and evaluation guidelines for underground piping for the Department of Energy (DOE) High-Level-Waste (HLW) Facilities. The underground piping includes both single and double containment steel pipes and concrete pipes with steel lining, with particular emphasis on the double containment piping. The design and evaluation guidelines presented in this paper follow the generally accepted beam-on-elastic-foundation analysis principle and the inertial response calculation method, respectively, for piping directly in contact with the soil or contained in a jacket. A standard analysis procedure is described along with the discussion of factors deemed to be significant for the design of the underground piping. The following key considerations are addressed: the design feature and safety requirements for the inner (core) pipe and the outer pipe; the effect of soil strain and wave passage; assimilation of the necessary seismic and soil data; inertial response calculation for the inner pipe; determination of support anchor movement loads; combination of design loads; and code comparison. Specifications and justifications of the key parameters used, stress components to be calculated and the allowable stress and strain limits for code evaluation are presented.

Lin, Chi-Wen [Consultant, Martinez, CA (United States); Antaki, G. [Westinghouse Savannah River Co., Aiken, SC (United States); Bandyopadhyay, K. [Brookhaven National Lab., Upton, NY (United States); Bush, S.H. [Review & Synthesis Association, Richland, WA (United States); Costantino, C. [City Univ. of New York, New York, NY (United States); Kennedy, R. [RPK Structural Mechanics, Yorba Linda, CA (United States). Consultant

1995-05-01T23:59:59.000Z

73

FINAL REPORT INTEGRATED DM1200 MELTER TESTING USING AZ 102 AND C 106/AY-102 HLW SIMULANTS: HLW SIMULANT VERIFICATION VSL-05R5800-1 REV 0 6/27/05  

Science Conference Proceedings (OSTI)

The principal objectives of the DM1200 melter tests were to determine the effects of feed rheology, feed solid content, and bubbler configuration on glass production rate and off-gas system performance while processing the HLW AZ-101 and C-106/AY-102 feed compositions; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components, as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and post test inspections of system components. The specific objectives (including test success criteria) of this testing, along with how each objective was met, are outlined in a table. The data provided in this Final Report address the impacts of HLW melter feed rheology on melter throughput and validation of the simulated HLW melter feeds. The primary purpose of this testing is to further validate/verify the HLW melter simulants that have been used for previous melter testing and to support their continued use in developing melter and off-gas related processing information for the Project. The primary simulant property in question is rheology. Simulants and melter feeds used in all previous melter tests were produced by direct addition of chemicals; these feed tend to be less viscous than rheological the upper-bound feeds made from actual wastes. Data provided here compare melter processing for the melter feed used in all previous DM100 and DM1200 tests (nominal melter feed) with feed adjusted by the feed vendor (NOAH Technologies) to be more viscous, thereby simulating more closely the upperbounding feed produced from actual waste. This report provides results of tests that are described in the Test Plan for this work. The Test Plan is responsive to one of several test objectives covered in the WTP Test Specification for this work; consequently, only part of the scope described in the Test Specification was addressed in this particular Test Plan. For the purpose of comparison, the tests reported here were performed with AZ-102 and C-106/AY-102 HLW simulants and glass compositions that are essentially the same as those used for recent DM1200 tests. One exception was the use of an alternate, higher-waste-loading C-106/AY-102 glass composition that was used in previous DM100 tests to further evaluate the performance of the optimized bubbler configuration.

KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D'ANGELO NA; BRANDYS M; KOT WK; PEGG IL

2011-12-29T23:59:59.000Z

74

Development Of High Waste-Loading HLW Glasses For High Bismuth Phosphate Wastes, VSL-12R2550-1, Rev 0  

SciTech Connect

This report presents results from tests with new glass formulations that have been developed for several high Bi-P HLW compositions that are expected to be processed at the WTP that have not been tested previously. WTP HLW feed compositions were reviewed to select waste batches that are high in Bi-P and that are reasonably distinct from the Bi-limited waste that has been tested previously. Three such high Bi-P HLW compositions were selected for this work. The focus of the present work was to determine whether the same type of issues as seen in previous work with high-Bi HLW will be seen in HLW with different concentrations of Bi, P and Cr and also whether similar glass formulation development approaches would be successful in mitigating these issues. New glass compositions were developed for each of the three representative Bi-P HLW wastes and characterized with respect to key processing and product quality properties and, in particular, those relating to crystallization and foaming tendency.

Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Gan, Hao [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States)

2012-12-13T23:59:59.000Z

75

Fluctuations of intensive quantities in statistical thermodynamics  

E-Print Network (OSTI)

In phenomenological thermodynamics, the canonical coordinates of a physical system split in pairs with each pair consisting of an extensive quantity and an intensive one. In the present paper, the quasi-thermodynamic fluctuation theory of a model system of a large number of oscillators is extended to statistical thermodynamics based on the idea to perceive the fluctuations of intensive variables as the fluctuations of specific extensive ones in a "thermodynamically dual" system. The extension is motivated by the symmetry of the problem in the context of an analogy with quantum mechanics which is stated in terms of a generalized Pauli problem for the thermodynamic fluctuations. The doubled Boltzmann constant divided by the number of particles plays a similar role to the Planck constant.

Artur E. Ruuge

2013-09-17T23:59:59.000Z

76

Final Report - Melt Rate Enhancement for High Aluminum HLW Glass Formulation, VSL-08R1360-1, Rev. 0, dated 12/19/08  

SciTech Connect

The principal objective of the work reported here was to develop and identify HLW glass compositions that maximize waste processing rates for the aluminum limted waste composition specified by ORP while maintaining high waste loadings and acceptable glass properties. This was accomplished through a combination of crucible-scale tests, confirmation tests on the DM100 melter system, and demonstration at pilot scale (DM1200). The DM100-BL unit was selected for these tests since it was used previously with the HLW waste streams evaluated in this study, was used for tests on HLW glass compositions to support subsequent tests on the HLW Pilot Melter, conduct tests to determine the effect of various glass properties (viscosity and conductivity) and oxide concentrations on glass production rates with HLW feed streams, and to assess the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition. The same melter was selected for the present tests in order to maintain comparisons between the previously collected data. These tests provide information on melter processing characteristics and off-gas data, including formation of secondary phases and partitioning. Once DM100 tests were completed, one of the compositions was selected for further testing on the DM1200; the DM1200 system has been used for processing a variety of simulated Hanford waste streams. Tests on the larger melter provide processing data at one third of the scale of the actual WTP HLW melter and, therefore, provide a more accurate and reliable assessment of production rates and potential processing issues. The work focused on maximizing waste processing rates for high aluminum HLW compositions. In view of the diversity of forms of aluminum in the Hanford tanks, tests were also conducted on the DM100 to determine the effect of changes in the form of aluminum on feed properties and production rate. In addition, the work evaluated the effect on production rate of modest increases in melter operating temperature. Glass composition development was based on one of the HLW waste compositions specified by ORP that has a high concentration of aluminum. Small-scale tests were used to provide an initial screening of various glass formulations with respect to melt rates; more definitive screening was provided by the subsequent DM100 tests. Glass properties evaluated included: viscosity, electrical conductivity, crystallinity, gross glass phase separation and the 7- day Product Consistency Test (ASTM-1285). Glass property limits were based upon the reference properties for the WTP HLW melter. However, the WTP crystallinity limit (< 1 vol% at 950oC) was relaxed slightly as a waste loading constraint for the crucible melts.

Kruger, Albert A.; Pegg, I. L.; Chaudhuri, M.; Gong, W.; Gan, H.; Matlack, K. S.; Bardakci, T.; Kot, W.

2013-11-13T23:59:59.000Z

77

Dissolution of ORNL HLW sludge and partitioning of the actinides using the TRUEX process  

SciTech Connect

Experiments were conducted to evaluate the transuranium extraction (TRUEX) process for partitioning actinides from actual dissolved high-level radioactive waste (HLW) sludge. Samples of sludge from melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign the rinsed sludge was leached in nitric acid, and about 50% of the dry mass of the sludge was dissolved. The resulting solution contained total metal concentrations of {approximately} 1.8 M with a nitric acid concentration of 2.9 M. In the other campaign the sludge was neutralized with nitric acid to destroy the carbonates, then leached with 2.6 M NaOH for {approximately} 6 h before rinsing with the mild caustic. The sludge was then leached in nitric acid, and about 80% of the sludge dissolved. The resulting solution contained total metal concentrations of {approximately} 0.6 M with a nitric acid concentration of 1.7 M. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th and U species removed from aqueous solution in only one extraction stage were > 95% and were close to the values predicted by the GTM. Mercury was also found to be strongly extracted, with a one-stage removal of > 92%. In one test, vanadium appeared to be moderately extracted.

Spencer, B.B.; Egan, B.Z.; Beahm, E.C.; Chase, C.W.; Dillow, T.A.

1997-12-01T23:59:59.000Z

78

Measuring and Predicting Fission Product Noble Metals in SRS HLW Sludges  

DOE Green Energy (OSTI)

The noble metals Ru, Rh, Pd, and Ag were produced in the Savannah River Site (SRS) reactors as products of the fission of U-235. Consequently they are in the High Level Waste (HLW) sludges that are currently being immobilized into a borosilicate glass in the Defense Waste Processing Facility (DWPF). The noble metals are a concern in the DWPF because they catalyze the decomposition of formic acid used in the process to produce the flammable gas hydrogen. As the concentration of these noble metals in the sludge increases, more hydrogen will be produced when this sludge is processed. In the SRS Tank Farm it takes approximately two years to prepare a sludge batch for processing in the DWPF. This length of time is necessary to mix the appropriate sludges, blend them to form a sludge batch and then wash it to enable processing in the DWPF. This means that the exact composition of a sludge batch is not known for {approx}two years. During this time, studies with simulated nonradioactive sludges must be performed to determine the desired DWPF processing parameters for the new sludge batch. Consequently, prediction of the noble metal concentrations is desirable to prepare appropriate simulated sludges for studies of the DWPF process for that sludge batch. These studies give a measure of the amount of hydrogen that will be produced when that sludge batch is processed. This report describes in detail the measurement of these noble metal concentrations in sludges and a way to predict their concentrations from an estimate of the lanthanum concentration in the sludge. Results for two sludges are presented in this report. These are Sludge Batch 3 (SB3) currently being processed by the DWPF and a sample of unwashed sludge from Tank 11 that will be part of Sludge Batch 4. The concentrations of the noble metals in HLW sludges are measured by using mass spectroscopy to determine concentrations of the isotopes that comprise each noble metal. For example, the noble metal Ru is comprised of isotopes with masses 101, 102, and 104. The element Rh has a single isotope with mass 103. The element Pd is comprised of five isotopes. These are at masses 105-108 and mass 110. As does Rh, Ag has only one isotope. This is at mass 109. However, results in this report show that the Ag concentration in the two samples was due to natural Ag being in the samples. Natural Ag has masses at 107 and 109. The Ag-107 interferes with the measurement of Pd-107. This Ag was used in one of the processes at SRS. The results also show that natural Cd is in the two samples. Cadmium has isotopes at masses 106, 108 and 110, thus it interferes with the analysis of the Pd isotopes at these masses. Cadmium was also used in one of the processes at SRS. However, the concentrations of the Pd isotopes at masses 106, 107, 108 and 110 could be calculated using the fission yields for the Pd isotopes, and the measured concentration of Pd at mass 105 where there is no Ag or Cd interference. Based on the measurements of the concentrations of the isotopes of each noble metal, the total concentration of that noble metal can be determined by summing the concentrations of the individual isotopes. The results in this report show that the relative concentrations of the isotopes of Ru and Rh are in proportion to their yields from the fission of U-235 in the reactors. These results were expected since these elements are very insoluble in caustic and thus are primarily in the sludge tanks rather then the salt tanks of the SRS Tank Farm. The relative concentration of Pd is somewhat lower than that based on the relative fission yields of its five isotopes. This indicates that some of the Pd is in the salt tanks rather than the sludge tanks of the Tank Farm. The concentrations of the noble metals were predicted using the High Level Waste Characterization System (WCS) at SRS. This system keeps record of the inventory of the major compounds and select radionuclides that are in each of the SRS HLW tanks. Using this system, the Closure Business Unit (CBU) can predict the major composition of a sludge ba

Bibler, N

2005-04-05T23:59:59.000Z

79

Category 3 threshold quantities for hazard categorization of nonreactor facilities  

Science Conference Proceedings (OSTI)

This document provides the information necessary to determine Hazard Category 3 threshold quantities for those isotopes of interest not listed in WHC-CM-4-46, Section 4, Table 1.''Threshold Quantities.''

Mandigo, R.L.

1996-02-13T23:59:59.000Z

80

Electrical Quantities Programs/Projects in Quantum Electrical ...  

Science Conference Proceedings (OSTI)

Electrical Quantities Programs/Projects in Quantum Electrical Metrology. Electric Power Metrology and the Smart Grid. Contact. ...

2011-10-03T23:59:59.000Z

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

Dose Calculations for the Co-Disposal WP-of HLW-Glass and the Triga SNF  

SciTech Connect

This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Operations (WPO). The purpose of this calculation is to determine the surface dose rates of a codisposal waste package (WP) containing a centrally located Department of Energy (DOE) standardized 18-in. spent nuclear fuel (SNF) canister, loaded with the TRIGA (Training, Research, Isotopes, General Atomics) SNF. This canister is surrounded by five 3-m long canisters, loaded with Savannah River Site (SRS) high-level waste (HLW) glass. The results are to support the WP design and radiological analyses.

G. Radulescu

1999-08-02T23:59:59.000Z

82

FY-97 operations of the pilot-scale glass melter to vitrify simulated ICPP high activity sodium-bearing waste  

SciTech Connect

A 3.5 liter refractory-lined joule-heated glass melter was built to test the applicability of electric melting to vitrify simulated high activity waste (HAW). The HAW streams result from dissolution and separation of Idaho Chemical Processing Plant (ICPP) calcines and/or radioactive liquid waste. Pilot scale melter operations will establish selection criteria needed to evaluate the application of joule heating to immobilize ICPP high activity waste streams. The melter was fabricated with K-3 refractory walls and Inconel 690 electrodes. It is designed to be continuously operated at 1,150 C with a maximum glass output rate of 10 lbs/hr. The first set of tests were completed using surrogate HAW-sodium bearing waste (SBW). The melter operated for 57 hours and was shut down due to excessive melt temperatures resulting in low glass viscosity (< 30 Poise). Due to the high melt temperature and low viscosity the molten glass breached the melt chamber. The melter has been dismantled and examined to identify required process improvement areas and successes of the first melter run. The melter has been redesigned and is currently being fabricated for the second run, which is scheduled to begin in December 1997.

Musick, C.A.

1997-11-01T23:59:59.000Z

83

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

Science Conference Proceedings (OSTI)

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

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

2006-06-30T23:59:59.000Z

84

FINAL REPORT DM1200 TESTS WITH AZ 101 HLW SIMULANTS VSL-03R3800-4 REV 0 2/17/04  

SciTech Connect

This report documents melter and off-gas performance results obtained on the DM 1200 HLW Pilot Melter during processing of simulated HLW AZ-101 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW AZ-101 feed; determine the effect of bubbling rate and feed solids content on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post-test inspections of system components. The test objectives (including test success criteria), along with how they were met, are outlined in a table.

KRUGER AA; MATLACK KS; BARDAKCI T; D'ANGELO NA; GONG W; KOT WK; PEGG IL

2011-12-29T23:59:59.000Z

85

BENEFITS OF VIBRATION ANALYSIS FOR DEVELOPMENT OF EQUIPMENT IN HLW TANKS - 12341  

Science Conference Proceedings (OSTI)

Vibration analyses of equipment intended for use in the Savannah River Site (SRS) radioactive liquid waste storage tanks are performed during pre-deployment testing and has been demonstrated to be effective in reducing the life-cycle costs of the equipment. Benefits of using vibration analysis to identify rotating machinery problems prior to deployment in radioactive service will be presented in this paper. Problems encountered at SRS and actions to correct or lessen the severity of the problem are discussed. In short, multi-million dollar cost saving have been realized at SRS as a direct result of vibration analysis on existing equipment. Vibration analysis of equipment prior to installation can potentially reduce inservice failures, and increases reliability. High-level radioactive waste is currently stored in underground carbon steel waste tanks at the United States Department of Energy (DOE) Savannah River Site and at the Hanford Site, WA. Various types of rotating machinery (pumps and separations equipment) are used to manage and retrieve the tank contents. Installation, maintenance, and repair of these pumps and other equipment are expensive. In fact, costs to remove and replace a single pump can be as high as a half million dollars due to requirements for radioactive containment. Problems that lead to in-service maintenance and/or equipment replacement can quickly exceed the initial investment, increase radiological exposure, generate additional waste, and risk contamination of personnel and the work environment. Several different types of equipment are considered in this paper, but pumps provide an initial example for the use of vibration analysis. Long-shaft (45 foot long) and short-shaft (5-10 feet long) equipment arrangements are used for 25-350 horsepower slurry mixing and transfer pumps in the SRS HLW tanks. Each pump has a unique design, operating characteristics and associated costs, sometimes exceeding a million dollars. Vibration data are routinely collected during pre-installation tests and screened for: Critical speeds or resonance, Imbalance of rotating parts, Shaft misalignment, Fluid whirl or lubrication break down, Bearing damages, and Other component abnormalities. Examples of previous changes in operating parameters and fabrication tolerances and extension of equipment life resulting from the SRS vibration analysis program include: (1) Limiting operational speeds for some pumps to extend service life without design or part changes; (2) Modifying manufacturing methods (tightening tolerances) for impellers on slurry mixing pumps based on vibration data that indicated hydraulic imbalance; (3) Identifying rolling element mounting defects and replacing those components in pump seals before installation; and (4) Identifying the need for bearing design modification for SRS long-shaft mixing pump designs to eliminate fluid whirl and critical speeds which significantly increased the equipment service life. In addition, vibration analyses and related analyses have been used during new equipment scale-up tests to identify the need for design improvements for full-scale operation / deployment of the equipment in the full size tanks. For example, vibration analyses were recently included in the rotary micro-filtration scale-up test program at SRNL.

Stefanko, D.; Herbert, J.

2012-01-10T23:59:59.000Z

86

The integrated economic production quantity model for inventory and quality.  

E-Print Network (OSTI)

??Determining the optimal production lot sizing has been widely used by the classical economic production quantity (EPQ) model. However, the analysis for finding an EPQ (more)

Ittharat, Tharat

2004-01-01T23:59:59.000Z

87

A Ceramic membrane to Recycle Caustic  

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

A A Ceramic Membrane to Recycle Caustic in Low-Activity Waste Stream Processing The Office of Waste Processing is sponsoring an R&D project with Ceramatec, Inc. to develop a ceramic membrane capable of separating sodium from the Hanford Low Activity Waste (LAW) stream. The Hanford High-Level Waste (HLW) tanks must be maintained in a caustic environment to inhibit corrosion. Consequently, they contain large quantities of NaOH. Ultimately the HLW will be retrieved, separated into HLW and LAW streams, with both streams being vitrified at the Waste Treatment Plant (WTP). Prior to processing, additional NaOH will be added to the LAW stream to solubilize the alumina, preventing alumina precipitation, but further increasing the NaOH quantity. This project's goal is to separate the sodium from the LAW stream prior to vitrification which will allow the NaOH to be recycled and further

88

Some Intensive and Extensive Quantities in High-Energy Collisions  

E-Print Network (OSTI)

We review the evolution of some statistical and thermodynamical quantities measured in difference sizes of high-energy collisions at different energies. We differentiate between intensive and extensive quantities and discuss the importance of their distinguishability in characterizing possible critical phenomena of nuclear collisions at various energies with different initial conditions.

Tawfik, A

2013-01-01T23:59:59.000Z

89

Advances in Glass Formulations for Hanford High-Alumimum, High-Iron and Enhanced Sulphate Management in HLW Streams - 13000  

SciTech Connect

The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet Hanford Tank Waste Treatment and Immobilization Plant (WTP) Contract terms. The WTP?s overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulphur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings and higher throughput efficiencies. Results of this work have demonstrated the feasibility of increases in waste loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. In view of the importance of aluminum limited waste streams at Hanford (and also Savannah River), the ability to achieve high waste loadings without adversely impacting melt rates has the potential for enormous cost savings from reductions in canister count and the potential for schedule acceleration. Consequently, the potential return on the investment made in the development of these enhancements is extremely favorable. Glass composition development for one of the latest Hanford HLW projected compositions with sulphate concentrations high enough to limit waste loading have been successfully tested and show tolerance for previously unreported tolerance for sulphate. Though a significant increase in waste loading for high-iron wastes has been achieved, the magnitude of the increase is not as substantial as those achieved for high-aluminum, high-chromium, high-bismuth or sulphur. Waste processing rate increases for high-iron streams as a combined effect of higher waste loadings and higher melt rates resulting from new formulations have been achieved.

Kruger, Albert A.

2013-01-16T23:59:59.000Z

90

ARM - Evaluation Product - Critical soil quantities for describing land  

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

ProductsCritical soil quantities for describing land ProductsCritical soil quantities for describing land properties Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Evaluation Product : Critical soil quantities for describing land properties 1994.01.01 - 2012.12.31 Site(s) SGP General Description The ARMBELAND is a subset of the ARM Best Estimate (ARMBE) products for supporting community land-atmospheric research and land model developments. It contains several critical soil quantities that ARM has been measuring for many years for describing land properties. The quantities in ARMBE-Land are averaged over one hour time interval, consistent with other ARMBE datasets. It is recommended to use with other ARMBE data products such as ARMBECLDRAD (cloud and radiative fluxes) and ARMBEATM (surface

91

Vitrification of DOE Problematic Wastes Using Iron Phosphate Glasses  

Science Conference Proceedings (OSTI)

Abstract Scope, This work is to formulate and optimize iron phosphate glass compositions which are suitable for vitrifying several specified Hanford HLW and ...

92

Analysis of Surface Leaching Processes in Vitrified High-Level Nuclear Wastes Using In-Situ Raman Imaging and Atomistic Modeling - Final Report  

SciTech Connect

The in situ analysis of surface conditions of vitrified nuclear wastes can provide an important check of the burial status of radioactive objects without risk of radiation exposure. Raman spectroscopy was initially chosen as the most promising method for testing the surface conditions of glasses undergoing chemical corrosion, and was used extensively during the first year. However, it was determined that infrared reflection spectroscopy was better suited to this particular need and was used for the remaining two years to investigate the surface corrosion behavior of model silicate glasses for extension to nuclear waste glasses. The developed methodology is consistent with the known theory of optical propagation of dielectric media and uses the Kramers-Kronig formalism. The results show that it is possible to study the corrosion of glass by analyzing the glass surface using reflection fast Fourier infrared measurements and the newly developed ''dispersion analysis method.'' The data show how this analysis can be used to monitor the corrosion behavior of vitrified waste glasses over extended periods of storage.

Simmons, Joseph H.

2001-04-24T23:59:59.000Z

93

Hanford high level waste (HLW) tank mixer pump safe operating envelope reliability assessment  

DOE Green Energy (OSTI)

The US Department of Energy and its contractor, Westinghouse Corp., are responsible for the management and safe storage of waste accumulated from processing defense reactor irradiated fuels for plutonium recovery at the Hanford Site. These wastes, which consist of liquids and precipitated solids, are stored in underground storage tanks pending final disposition. Currently, 23 waste tanks have been placed on a safety watch list because of their potential for generating, storing, and periodically releasing various quantities of hydrogen and other gases. Tank 101-SY in the Hanford SY Tank Farm has been found to release hydrogen concentrations greater than the lower flammable limit (LFL) during periodic gas release events. In the unlikely event that an ignition source is present during a hydrogen release, a hydrogen burn could occur with a potential to release nuclear waste materials. To mitigate the periodic gas releases occurring from Tank 101-SY, a large mixer pump currently is being installed in the tank to promote a sustained release of hydrogen gas to the tank dome space. An extensive safety analysis (SA) effort was undertaken and documented to ensure the safe operation of the mixer pump after it is installed in Tank 101-SY.1 The SA identified a need for detailed operating, alarm, and abort limits to ensure that analyzed safety limits were not exceeded during pump operations.

Fischer, S.R. [Los Alamos National Lab., NM (United States); Clark, J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)

1993-10-01T23:59:59.000Z

94

South Dakota Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) South Dakota Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 1,846 1,947 2,558 2,231 3,431 3,920 4,369 1990's 881 93 1,006 854 1,000 848 0 687 772 702 2000's 648 563 531 550 531 446 455 422 1,099 NA 2010's NA - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 1/7/2014 Next Release Date: 1/31/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead Value South Dakota Natural Gas Wellhead Value and Marketed Production

95

EM's Indefinite Delivery/Indefinite Quantity Cleanup Contracts |  

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

EM's Indefinite Delivery/Indefinite Quantity Cleanup Contracts EM's Indefinite Delivery/Indefinite Quantity Cleanup Contracts EM's Indefinite Delivery/Indefinite Quantity Cleanup Contracts The Office of Environmental Management (EM) has 23 Indefinite Delivery/Indefinite Quantity (IDIQ) contracts to provide cleanup services at EM sites across the United States. The scope of work of the IDIQ contracts includes: environmental remediation deactivation, decommissioning, demolition and removal of contaminated facilities waste management regulatory compliance These nationwide, multiple-award IDIQ contracts allow EM sites to place timely, competitive and cost-effective task orders for environmental services with either large or small businesses, as determined by the complexity of the requirements. Twelve of the IDIQ contracts were awarded

96

Indiana Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Indiana Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 135 394 367 365 217 412 416 1990's 399 232 174 192 107 249 360 526 615 855 2000's 899 1,064 1,309 1,464 3,401 3,135 2,921 3,606 4,701 4,927 2010's 6,802 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead Value Indiana Natural Gas Wellhead Value and Marketed Production

97

Integrated quality and quantity modeling of a production line  

E-Print Network (OSTI)

The interaction of quantity and quality performance in a factory is clearly of great economic importance. However, there is very little quantitative analytical literature in this area. This thesis is an essential early ...

Kim, Jongyoon, 1974-

2005-01-01T23:59:59.000Z

98

Electricity supply chain coordination based on quantity discount contracts  

Science Conference Proceedings (OSTI)

Electricity supply chain coordination mechanism from the fuel supply, power generation and transmission to electricity consumption has become an important research topic to ease electric coal supply conflicts. In this paper, based on the quantity discount ...

Yu Dai; Hongming Yang; Jiajie Wu

2009-02-01T23:59:59.000Z

99

Implications of MoninObukhov Similarity Theory for Scalar Quantities  

Science Conference Proceedings (OSTI)

Monin-Obukhov similarity theory of surface-layer turbulence has been extended to include all scalar quantities. The tenets of this theory, as it is presently practiced, are followed to their logical conclusions, which produce some novel results. ...

Reginald J. Hill

1989-07-01T23:59:59.000Z

100

Final Report - Effects of High Spinel and Chromium Oxide Crystal Contents on Simulated HLW Vitrification in DM100 Melter Tests, VSL-09R1520-1, Rev. 0, dated 6/22/09  

SciTech Connect

The principal objective of the work was to evaluate the effects of spinel and chromium oxide particles on WTP HLW melter operations and potential impacts on melter life. This was accomplished through a combination of crucible-scale tests, settling and rheological tests, and tests on the DM100 melter system. Crucible testing was designed to develop and identify HLW glass compositions with high waste loadings that exhibit formation of crystalline spinel and/or chromium oxide phases up to relatively high crystal contents (i.e., > 1 vol%). Characterization of crystal settling and the effects on melt rheology was performed on the HLW glass formulations. Appropriate candidate HLW glass formulations were selected, based on characterization results, to support subsequent melter tests. In the present work, crucible melts were formulated that exhibit up to about 4.4 vol% crystallization.

Kruger, Albert A.; Matlack, K. S.; Kot, W.; Pegg, I. L.; Chaudhuri, M.; Lutze, W.

2013-11-13T23:59:59.000Z

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

APPLICATION OF MECHANICAL ACTIVATION TO PRODUCTION OF PYROCHLORE CERAMIC CONTAINING SIMULATED RARE-EARTH ACTINIDE FRACTION OF HLW  

SciTech Connect

Samples of zirconate pyrochlore ceramic (REE)2(Zr,U)2O7 (REE = La-Gd) containing simulated REE-An fraction of HLW were synthesized by two routes: (1) conventional cold compaction of oxide mixtures in pellets under pressure of 200 MPa and sintering of the pellets at 1550 C for 24 hours; and (2) using preliminary mechanical activation of oxide powders in a linear inductive rotator (LIV-0.5E) and a planetary mill - activator with hydrostatic yokes (AGO-2U) for 5 or 10 min. All the samples sintered at 1550 C were monolithic and dense with high mechanical integrity. As follows from X-ray diffraction (XRD) data, the ceramic sample produced without mechanical activation is composed of pyrochlore as major phase but contains also minor unreacted oxides. The samples prepared from pre-activated mixtures are composed of the pyrochlore structure phase only. Scanning electron microscopy (SEM) data also show higher structural and compositional homogeneity of the samples prepared from mechanically activated batches. The samples produced from oxide mixtures mechanically activated in the LIV for 10 min were slightly contaminated with iron resulting in formation of minor perovskite structure phase not detected by XRD but seen on SEM-images of the samples. Comparison of the samples prepared from non-activated and activated batches showed higher density, lower open porosity, water uptake, and elemental leaching for the samples fabricated from mechanically activated oxide mixtures.

Stefanovsky, S.V.; Kirjanova, O.I.; Chizhevskaya, S.V.; Yudintsev, S.V.; Nikonov, B.S.

2003-02-27T23:59:59.000Z

102

Low Level Waste Disposition - Quantity and Inventory | Department of  

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

Low Level Waste Disposition - Quantity and Inventory Low Level Waste Disposition - Quantity and Inventory Low Level Waste Disposition - Quantity and Inventory This study has been prepared by the Used Fuel Disposition (UFD) campaign of the Fuel Cycle Research and Development (FCR&D) program. The purpose of this study is to provide an estimate of the volume of low level waste resulting from a variety of commercial fuel cycle alternatives in order to support subsequent system-level evaluations of disposal system performance. This study provides an estimate of Class A/B/C low level waste (LLW), greater than Class C (GTCC) waste, mixed LLW and mixed GTCC waste generated from the following initial set of fuel cycles and recycling processes: 1. Operations at a geologic repository based upon a once through light

103

Low Level Waste Disposition - Quantity and Inventory | Department of  

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

Low Level Waste Disposition - Quantity and Inventory Low Level Waste Disposition - Quantity and Inventory Low Level Waste Disposition - Quantity and Inventory This study has been prepared by the Used Fuel Disposition (UFD) campaign of the Fuel Cycle Research and Development (FCR&D) program. The purpose of this study is to provide an estimate of the volume of low level waste resulting from a variety of commercial fuel cycle alternatives in order to support subsequent system-level evaluations of disposal system performance. This study provides an estimate of Class A/B/C low level waste (LLW), greater than Class C (GTCC) waste, mixed LLW and mixed GTCC waste generated from the following initial set of fuel cycles and recycling processes: 1. Operations at a geologic repository based upon a once through light

104

TESTS WITH HIGH-BISMUTH HLW GLASSES FINAL REPORT VSL-10R1780-1 REV 0 12/13/10  

Science Conference Proceedings (OSTI)

This Final Report describes the testing of glass formulations developed for Hanford High Level Waste (HLW) containing high concentrations of bismuth. In previous work on high-bismuth HLW streams specified by the Office of River Protection (ORP), fully compliant, high waste loading compositions were developed and subjected to melter testing on the DM100 vitrification system. However, during heat treatment according to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW canister centerline cooling (CCC) curves, crucible melts of the high-bismuth glasses were observed to foam. Clearly, such an occurrence during cooling of actual HLW canisters would be highly undesirable. Accordingly, the present work involves larger-scale testing to determine whether this effect occurs under more prototypical conditions, as well as crucible-scale tests to determine the causes and potentially remediate the observed foaming behavior. The work included preparation and characterization of crucible melts designed to determine the underlying causes of the foaming behavior as well as to assess potential mitigation strategies. Testing was also conducted on the DM1200 HLW Pilot melter with a composition previously tested on the DM100 and shown to foam during crucible-scale CCC heat treatment. The DM1200 tests evaluated foaming of glasses over a range of bismuth concentrations poured into temperature-controlled, 55-gallon drums which have a diameter that is close to that of the full-scale WTP HLW canisters. In addition, the DM1200 tests provided the first large-scale melter test data on high-bismuth WTP HLW compositions, including information on processing rates, cold cap behavior and off-gas characteristics, and data from this waste composition on the prototypical DM1200 off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for ORP on the same waste composition. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. The present glass formulation and melter testing work was aimed at one of the four waste streams previously specified by the Office of River Protection (ORP). Such testing supports the ORP basis for projection of the amount of Immobilized High Level Waste (IHLW) to be produced at Hanford and evaluation of the likely potential for future enhancements of the WTP over and above the present well-developed baseline. It should be noted that the compositions of the four ORP-specified waste streams differ significantly from those of the feed tanks (AZ-101, AZ-102, C-16/AY-102, and C-104/AY-101) that have been the focus of the extensive technology development and design work performed for the WTP baseline. In this regard, the work on the ORP-specified compositions is complementary to and necessarily of a more exploratory nature than the work in support of the current WTP baseline.

MATLACK KS; KRUGER AA; JOSEPH I; GAN H; KOT WK; CHAUDHURI M; MOHR RK; MCKEOWN DA; BARDAKEI T; GONG W; BUECCHELE AC; PEGG IL

2011-01-05T23:59:59.000Z

105

Michigan Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Michigan Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 138,910 144,537 131,855 127,287 146,996 146,145 155,988 1990's 106,193 189,497 190,637 199,746 216,268 238,203 245,740 305,950 278,076 277,364 2000's 296,556 275,036 274,476 236,987 259,681 261,112 NA NA 153,130 159,400 2010's 151,886 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

106

Arkansas Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Arkansas Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 78,097 75,575 86,552 68,206 42,688 102,046 42,226 1990's 99,456 83,864 85,177 122,596 24,326 180,117 76,671 71,449 61,012 54,382 2000's 55,057 16,901 161,871 166,329 183,299 190,533 193,491 269,886 446,551 680,613 2010's 936,600 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

107

West Virginia Quantity of Production Associated with Reported Wellhead  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) West Virginia Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 130,078 143,730 144,883 135,431 160,000 174,942 177,192 1990's 95,271 198,605 202,775 171,024 55,756 50,439 0 0 0 0 2000's 0 0 NA 0 NA NA NA NA NA NA 2010's NA - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead Value West Virginia Natural Gas Wellhead Value and Marketed

108

Kentucky Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Kentucky Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 46,720 61,518 73,126 80,195 70,125 44,725 72,417 1990's 75,333 78,904 79,690 86,966 73,081 74,754 81,435 79,547 81,868 76,770 2000's 81,545 81,723 88,259 87,609 94,259 92,795 95,320 95,437 114,116 NA 2010's 135,355 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 1/7/2014 Next Release Date: 1/31/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

109

Ohio Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Ohio Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 149,096 184,651 180,458 180,287 164,960 166,690 159,730 1990's 154,619 146,189 143,381 135,939 130,855 125,085 119,251 116,246 108,542 102,505 2000's 98,551 97,272 103,158 120,081 119,847 83,523 86,315 88,095 84,858 88,824 2010's 78,122 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

110

New York Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) New York Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 17,836 25,200 31,561 22,964 25,676 23,455 20,433 1990's 25,023 21,704 22,543 20,620 19,684 17,325 0 15,415 15,415 15,426 2000's 17,166 27,187 35,941 35,044 45,436 54,377 55,344 54,942 50,320 44,849 2010's 35,241 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

111

Virginia Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Virginia Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 4,342 8,928 15,041 15,427 19,223 18,424 17,935 1990's 14,283 14,906 24,734 37,840 50,259 49,818 0 0 0 0 2000's 0 0 0 0 NA NA NA NA NA NA 2010's NA - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead Value Virginia Natural Gas Wellhead Value and Marketed Production

112

Oregon Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Oregon Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 3 2,790 4,080 4,600 3,800 4,000 2,500 1990's 2,815 2,741 2,580 4,003 3,221 1,923 1,439 1,173 1,067 1,291 2000's 1,214 1,069 837 688 467 433 NA 390 751 751 2010's 1,376 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead Value Oregon Natural Gas Wellhead Value and Marketed Production

113

Alaska Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Alaska Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Alaska Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Alaska Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 211,569 211,579 222,637 304,841 271,120 228,284 192,760 1990's 191,798 200,557 206,259 224,786 201,891 227,797 193,278 191,017 192,982 186,727 2000's 189,896 197,735 200,871 199,616 413,667 502,887 494,323 368,344 337,359 397,077 2010's 316,546 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

114

Alabama Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Alabama Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Alabama Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Alabama Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 59,051 56,685 42,925 34,164 35,674 45,488 41,614 1990's 37,229 35,972 51,219 75,474 70,489 54,964 493,069 583,370 560,414 544,020 2000's 521,215 376,241 370,753 348,722 304,212 285,237 274,176 259,062 246,747 225,666 2010's 212,769 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 1/7/2014 Next Release Date: 1/31/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

115

Pennsylvania Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Pennsylvania Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 118,372 166,342 150,234 159,889 163,318 167,089 191,774 1990's 177,609 152,500 138,675 189,443 187,113 177,139 0 0 0 0 2000's 0 0 0 0 NA NA NA NA NA NA 2010's NA - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead Value Pennsylvania Natural Gas Wellhead Value and Marketed

116

Nebraska Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Nebraska Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 2,091 2,300 1,944 1,403 1,261 910 878 1990's 793 785 1,177 1,375 2,098 1,538 1,332 1,194 1,285 1,049 2000's 879 883 892 1,168 1,172 1,172 NA 1,555 3,082 2,908 2010's 2,231 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead Value Nebraska Natural Gas Wellhead Value and Marketed Production

117

Montana Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Montana Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 47,751 47,534 46,113 42,203 42,814 47,748 52,044 1990's 45,998 48,075 50,359 58,810 51,953 46,739 46,868 50,409 51,967 55,780 2000's 67,294 78,493 86,075 86,027 90,771 101,666 106,843 110,942 802,619 293,941 2010's 87,539 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

118

North Dakota Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) North Dakota Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 69,319 60,111 62,371 58,593 51,671 21,240 12,290 1990's 11,537 5,138 3,994 4,420 0 0 0 52,401 53,185 52,862 2000's 48,714 57,949 57,015 57,808 59,513 57,972 53,675 54,745 52,469 59,369 2010's 81,837 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 1/7/2014 Next Release Date: 1/31/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

119

California Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) California Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 282,639 343,079 361,739 329,366 346,720 327,399 283,509 1990's 275,738 211,841 195,515 76,381 199,649 263 37,823 219,216 264,810 382,715 2000's 323,864 328,778 309,399 293,691 276,520 274,817 278,933 268,016 263,107 241,916 2010's 251,559 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

120

Mississippi Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Mississippi Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 211,116 206,871 178,426 197,217 195,299 196,912 148,167 1990's 149,012 126,637 129,340 131,450 105,646 95,349 88,805 98,075 88,723 83,232 2000's 70,965 76,986 112,979 133,901 145,692 52,923 60,531 73,460 96,641 97,258 2010's 73,721 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

Colorado Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Colorado Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 139,820 143,552 126,037 163,684 164,557 191,544 216,737 1990's 242,997 271,159 314,105 388,016 441,343 511,513 559,473 637,375 696,321 705,477 2000's 735,332 800,712 819,205 989,678 1,058,383 1,106,993 1,170,819 1,280,638 1,436,203 1,409,172 2010's 1,548,576 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages: Quantity of Natural Gas Production Associated with Reported Wellhead

122

WHAT PHYSICAL QUANTITIES MAKE SENSE IN NONEQUILIBRIUM STATISTICAL MECHANICS?  

E-Print Network (OSTI)

WHAT PHYSICAL QUANTITIES MAKE SENSE IN NONEQUILIBRIUM STATISTICAL MECHANICS? by David Ruelle*. Abstract. Statistical mechanics away from equilibrium is in a formative stage, where general concepts;1 Introduction. Statistical mechanics, as seen by Boltzmann, is an attempt to understand the bulk properties

Ruelle, David

123

Wyoming Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Wyoming Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 395,656 447,615 416,565 352,858 407,863 471,095 623,915 1990's 690,356 711,799 765,254 63,667 14,283 12,449 27,821 719,933 1,004,020 1,079,375 2000's 1,240,038 1,359,868 1,533,724 1,561,322 1,724,725 1,729,760 1,811,992 1,916,238 2,116,818 2,239,778 2010's 2,318,486 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014 Referring Pages:

124

New Mexico Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) New Mexico Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 884,517 925,298 880,307 676,886 790,639 752,629 833,593 1990's 949,735 1,029,824 1,274,220 1,489,052 1,510,804 1,480,327 1,553,103 1,540,157 1,483,370 1,511,671 2000's 1,685,664 1,670,644 1,614,045 1,576,639 1,578,773 1,571,920 1,562,754 1,495,615 895,675 1,370,727 2010's 1,287,399 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014

125

Oklahoma Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Oklahoma Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 1,730,061 1,985,869 1,936,341 1,917,493 2,004,797 2,106,632 2,185,204 1990's 2,186,153 2,119,161 1,937,224 2,005,971 1,879,257 1,765,788 1,751,487 1,452,233 1,644,531 1,577,961 2000's 1,612,890 1,477,058 1,456,375 1,531,657 1,584,905 1,571,615 1,683,563 1,589,871 1,765,988 1,621,316 2010's 1,408,061 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014

126

Texas Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Texas Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 6,227,995 6,630,246 6,367,936 6,465,964 6,414,021 6,386,544 6,276,968 1990's 6,476,032 6,066,256 5,893,069 5,769,437 5,834,671 5,592,323 4,684,140 4,716,304 4,777,945 5,719,128 2000's 5,869,901 5,159,233 5,166,315 5,186,213 5,271,306 5,539,052 5,993,702 6,454,249 7,483,842 7,623,747 2010's 7,470,752 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 12/12/2013 Next Release Date: 1/7/2014

127

Louisiana Quantity of Production Associated with Reported Wellhead Value  

Gasoline and Diesel Fuel Update (EIA)

Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Louisiana Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 3,149,192 3,650,412 3,179,306 2,986,468 3,243,795 3,158,903 3,066,789 1990's 3,780,551 3,355,867 3,404,963 3,454,646 3,562,360 3,709,015 3,976,305 5,398,216 5,410,523 5,265,670 2000's 3,587,815 1,529,733 1,365,925 1,350,399 1,357,366 1,296,048 1,361,119 1,275,806 1,292,478 1,449,809 2010's 2,140,525 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date: 1/7/2014 Next Release Date: 1/31/2014

128

DATA SUMMARY REPORT SMALL SCALE MELTER TESTING OF HLW ALGORITHM GLASSES MATRIX1 TESTS VSL-07S1220-1 REV 0 7/25/07  

SciTech Connect

Eight tests using different HLW feeds were conducted on the DM100-BL to determine the effect of variations in glass properties and feed composition on processing rates and melter conditions (off-gas characteristics, glass processing, foaming, cold cap, etc.) at constant bubbling rate. In over seven hundred hours of testing, the property extremes of glass viscosity, electrical conductivity, and T{sub 1%}, as well as minimum and maximum concentrations of several major and minor glass components were evaluated using glass compositions that have been tested previously at the crucible scale. Other parameters evaluated with respect to glass processing properties were +/-15% batching errors in the addition of glass forming chemicals (GFCs) to the feed, and variation in the sources of boron and sodium used in the GFCs. Tests evaluating batching errors and GFC source employed variations on the HLW98-86 formulation (a glass composition formulated for HLW C-106/AY-102 waste and processed in several previous melter tests) in order to best isolate the effect of each test variable. These tests are outlined in a Test Plan that was prepared in response to the Test Specification for this work. The present report provides summary level data for all of the tests in the first test matrix (Matrix 1) in the Test Plan. Summary results from the remaining tests, investigating minimum and maximum concentrations of major and minor glass components employing variations on the HLW98-86 formulation and glasses generated by the HLW glass formulation algorithm, will be reported separately after those tests are completed. The test data summarized herein include glass production rates, the type and amount of feed used, a variety of measured melter parameters including temperatures and electrode power, feed sample analysis, measured glass properties, and gaseous emissions rates. More detailed information and analysis from the melter tests with complete emission chemistry, glass durability, and melter operating details will be provided in the final report. A summary of the tests that were conducted is provided in Table 1. Each of the seven tests was of nominally one hundred hours in duration. Test B was conducted in two equal segments: the first with nominal additives, and the second with the replacement of borax with a mixture of boric acid and soda ash to determine the effect of alternative OPC sources on production rates and processing characteristics. Interestingly, sugar additions were required near mid points of Tests W and Z to reduce excessive foaming that severely limited feed processing rates. The sugar additions were very effective in recovering manageable processing conditions, albeit over the relatively short remainder of the test duration. Tests W and Z employed the highest melt viscosities but not by a particularly wide margin. Other tests, which did not exhibit such foaming Issues, employed higher concentrations of manganese or iron or both. These results highlight the need for the development of protocols for the a priori determination of which HLW feeds will require sugar additions and the appropriate amounts of sugar to be added in order to control foaming (and maintain throughput) without over-reduction of the melt (which could lead to molten metal formation). In total, over 8,800 kg of feed was processed to produce over 3200 kg of glass. Steady-state processing rates were achieved, and no secondary sulfate phases were observed during any of the tests. Analysis was performed on samples of the glass product taken throughout the tests to verify composition and properties. Sampling and analysis was also performed on melter exhaust to determine the effect of the feed and glass changes on melter emissions.

KRUGER AA; MATLACK KS; PEGG IL

2011-12-29T23:59:59.000Z

129

High-level waste melter alternatives assessment report  

SciTech Connect

This document describes the Tank Waste Remediation System (TWRS) High-Level Waste (HLW) Program`s (hereafter referred to as HLW Program) Melter Candidate Assessment Activity performed in fiscal year (FY) 1994. The mission of the TWRS Program is to store, treat, and immobilize highly radioactive Hanford Site waste (current and future tank waste and encapsulated strontium and cesium isotopic sources) in an environmentally sound, safe, and cost-effective manner. The goal of the HLW Program is to immobilize the HLW fraction of pretreated tank waste into a vitrified product suitable for interim onsite storage and eventual offsite disposal at a geologic repository. Preparation of the encapsulated strontium and cesium isotopic sources for final disposal is also included in the HLW Program. As a result of trade studies performed in 1992 and 1993, processes planned for pretreatment of tank wastes were modified substantially because of increasing estimates of the quantity of high-level and transuranic tank waste remaining after pretreatment. This resulted in substantial increases in needed vitrification plant capacity compared to the capacity of original Hanford Waste Vitrification Plant (HWVP). The required capacity has not been finalized, but is expected to be four to eight times that of the HWVP design. The increased capacity requirements for the HLW vitrification plant`s melter prompted the assessment of candidate high-capacity HLW melter technologies to determine the most viable candidates and the required development and testing (D and T) focus required to select the Hanford Site HLW vitrification plant melter system. An assessment process was developed in early 1994. This document describes the assessment team, roles of team members, the phased assessment process and results, resulting recommendations, and the implementation strategy.

Calmus, R.B.

1995-02-01T23:59:59.000Z

130

INCONEL 690 CORROSION IN WTP (WASTE TREATMENT PLANT) HLW (HIGH LEVEL WASTE) GLASS MELTS RICH IN ALUMINUM & BISMUTH & CHROMIUM OR ALUMINUM/SODIUM  

SciTech Connect

Metal corrosion tests were conducted with four high waste loading non-Fe-limited HLW glass compositions. The results at 1150 C (the WTP nominal melter operating temperature) show corrosion performance for all four glasses that is comparable to that of other typical borosilicate waste glasses, including HLW glass compositions that have been developed for iron-limited WTP streams. Of the four glasses tested, the Bi-limited composition shows the greatest extent of corrosion, which may be related to its higher phosphorus content. Tests at higher suggest that a moderate elevation of the melter operating temperature (up to 1200 C) should not result in any significant increase in Inconel corrosion. However, corrosion rates did increase significantly at yet higher temperatures (1230 C). Very little difference was observed with and without the presence of an electric current density of 6 A/inch{sup 2}, which is the typical upper design limit for Inconel electrodes. The data show a roughly linear relationship between the thickness of the oxide scale on the coupon and the Cr-depletion depth, which is consistent with the chromium depletion providing the material source for scale growth. Analysis of the time dependence of the Cr depletion profiles measured at 1200 C suggests that diffusion of Cr in the Ni-based Inconel alloy controls the depletion depth of Cr inside the alloy. The diffusion coefficient derived from the experimental data agrees within one order of magnitude with the published diffusion coefficient data for Cr in Ni matrices; the difference is likely due to the contribution from faster grain boundary diffusion in the tested Inconel alloy. A simple diffusion model based on these data predicts that Inconel 690 alloy will suffer Cr depletion damage to a depth of about 1 cm over a five year service life at 1200 C in these glasses.

KRUGER AA; FENG Z; GAN H; PEGG IL

2009-11-05T23:59:59.000Z

131

A comparison of methods for representing sparsely sampled random quantities.  

SciTech Connect

This report discusses the treatment of uncertainties stemming from relatively few samples of random quantities. The importance of this topic extends beyond experimental data uncertainty to situations involving uncertainty in model calibration, validation, and prediction. With very sparse data samples it is not practical to have a goal of accurately estimating the underlying probability density function (PDF). Rather, a pragmatic goal is that the uncertainty representation should be conservative so as to bound a specified percentile range of the actual PDF, say the range between 0.025 and .975 percentiles, with reasonable reliability. A second, opposing objective is that the representation not be overly conservative; that it minimally over-estimate the desired percentile range of the actual PDF. The presence of the two opposing objectives makes the sparse-data uncertainty representation problem interesting and difficult. In this report, five uncertainty representation techniques are characterized for their performance on twenty-one test problems (over thousands of trials for each problem) according to these two opposing objectives and other performance measures. Two of the methods, statistical Tolerance Intervals and a kernel density approach specifically developed for handling sparse data, exhibit significantly better overall performance than the others.

Romero, Vicente Jose; Swiler, Laura Painton; Urbina, Angel; Mullins, Joshua

2013-09-01T23:59:59.000Z

132

Memo, "Incorporation of HLW Glass Shell V2.0 into the Flowsheets," to ED Lee, CCN: 184905, October 20, 2009  

Science Conference Proceedings (OSTI)

Efforts are being made to increase the efficiency and decrease the cost of vitrifying radioactive waste stored in tanks at the U.S. Department of Energy Hanford Site. The compositions of acceptable and processable high-level waste (HL W) glasses need to be optimized to minimize the waste-form volume and, hence, to reduce cost. A database of glass properties of waste glass and associated simulated waste glasses was collected and documented in PNNL 18501, Glass Property Data and Models for Estimating High-Level Waste Glass Volume and glass property models were curve-fitted to the glass compositions. A routine was developed that estimates HL W glass volumes using the following glass property models: II Nepheline, II One-Percent Crystal Temperature (T1%), II Viscosity (11) II Product Consistency Tests (PCT) for boron, sodium, and lithium, and II Liquidus Temperature (TL). The routine, commonly called the HL W Glass Shell, is presented in this document. In addition to the use of the glass property models, glass composition constraints and rules, as recommend in PNNL 18501 and in other documents (as referenced in this report) were incorporated. This new version of the HL W Glass Shell should generally estimate higher waste loading in the HL W glass than previous versions.

Gimpel, Rodney F.; Kruger, Albert A.

2013-12-18T23:59:59.000Z

133

FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF BUBBLER CONFIGURATIONS USING HLW AZ-101 SIMULANTS VSL-04R4800-4 REV 0 10/5/04  

SciTech Connect

This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 HLW simulants. The tests reported herein are a subset of six tests from a larger series of tests described in the Test Plan for the work; results from the other tests have been reported separately. The solids contents of the melter feeds were based on the WTP baseline value for the solids content of the feeds from pretreatment which changed during these tests from 20% to 15% undissolved solids resulting in tests conducted at two feed solids contents. Based on the results of earlier tests with single outlet 'J' bubblers, initial tests were performed with a total bubbling rate of 651 pm. The first set of tests (Tests 1A-1E) addressed the effects of skewing this total air flow rate back and forth between the two installed bubblers in comparison to a fixed equal division of flow between them. The second set of tests (2A-2D) addressed the effects of bubbler depth. Subsequently, as the location, type and number of bubbling outlets were varied, the optimum bubbling rate for each was determined. A third (3A-3C) and fourth (8A-8C) set of tests evaluated the effects of alternative bubbler designs with two gas outlets per bubbler instead of one by placing four bubblers in positions simulating multiple-outlet bubblers. Data from the simulated multiple outlet bubblers were used to design bubblers with two outlets for an additional set of tests (9A-9C). Test 9 was also used to determine the effect of small sugar additions to the feed on ruthenium volatility. Another set of tests (10A-10D) evaluated the effects on production rate of spiking the feed with chloride and sulfate. Variables held constant to the extent possible included melt temperature, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The feed rate was increased to the point that a constant, essentially complete, cold cap was achieved, which was used as an indicator of a maximized feed rate for each test. The first day of each test was used to build the cold cap and decrease the plenum temperature. The remainder of each test was split into two- to six-day segments, each with a different bubbling rate, bubbler orientation, or feed concentration of chloride and sulfur.

KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D'ANGELO NA; LUTZE W; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

2011-12-29T23:59:59.000Z

134

Table A26. Total Quantity of Purchased Energy Sources by Census...  

U.S. Energy Information Administration (EIA) Indexed Site

Total Quantity of Purchased Energy Sources by Census Region and" " Economic Characteristics of the Establishment, 1991" " (Estimates in Btu or Physical Units)"...

135

FCRD-USED-2010-000033, LLW Quantity and Inventory, FINAL R.....  

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

and Development Used Fuel Disposition Low Level Waste Disposition - Quantity and Inventory Prepared for U.S. Department of Energy Used Nuclear Fuel Robert H. Jones, SRS June...

136

FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF REDOX EFFECTS USING HLW AZ-101 AND C-106/AY-102 SIMULANTS VSL-04R4800-1 REV 0 5/6/  

SciTech Connect

This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 and C-106/AY-102 HLW simulants. The tests reported herein are a subset of three tests from a larger series of tests described in the Test Plan for the work; results from the remaining tests will be reported separately. Three nine day tests, one with AZ-101 and two with C-106/AY-102 feeds were conducted with variable amounts of added sugar to address the effects of redox. The test with AZ-101 included ruthenium spikes to also address the effects of redox on ruthenium volatility. One of tests addressed the effects of increased flow-sheet nitrate levels using C-106/AY-102 feeds. With high nitrate/nitrite feeds (such as WTP LAW feeds), reductants are required to prevent melt foaming and deleterious effects on glass production rates. Sugar is the baseline WTP reductant for this purpose. WTP HLW feeds typically have relatively low nitrate/nitrite content in comparison to the organic carbon content and, therefore, have typically not required sugar additions. However, HLW feed variability, particularly with respect to nitrate levels, may necessitate the use of sugar in some instances. The tests reported here investigate the effects of variable sugar additions to the melter feed as well as elevated nitrate levels in the waste. Variables held constant to the extent possible included melt temperature, bubbling rate, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW feeds with variable amounts of added sugar and increased nitrate levels; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and post test inspections of system components. The specific objectives (including test success criteria) of this testing, along with how each objective was met, are outlined in a table.

KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D'ANGELO NA; LUTZE W; BIZOT PM; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

2011-12-29T23:59:59.000Z

137

Table A23. Quantity of Purchased Electricity, Steam, and Natural Gas by Type  

U.S. Energy Information Administration (EIA) Indexed Site

3. Quantity of Purchased Electricity, Steam, and Natural Gas by Type" 3. Quantity of Purchased Electricity, Steam, and Natural Gas by Type" " of Supplier, Census Region, Industry Group, and Selected Industries, 1991" " (Estimates in Btu or Physical Units)" ,," Electricity",," Steam",," Natural Gas" ,," (Million kWh)",," (Billion Btu)",," (Billion cu ft)" ,," -------------------------",," -------------------------",," ---------------------------------------",,,"RSE" "SIC",,"Utility","Nonutility","Utility","Nonutility","Utility","Transmission","Other","Row"

138

Table A27. Quantity of Purchased Electricity, Steam, and Natural Gas by Type  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Purchased Electricity, Steam, and Natural Gas by Type" Quantity of Purchased Electricity, Steam, and Natural Gas by Type" " of Supplier, Census Region, and Economic Characteristics of the Establishment," 1991 " (Estimates in Btu or Physical Units)" " "," Electricity",," Steam",," Natural Gas" ," (Million (kWh)",," (Billion Btu)",," (Billion cu ft)" ," -----------------------",," -----------------------",," ------------------------------------",,,"RSE" ,"Utility","Nonutility","Utility","Nonutility","Utility","Transmission","Other","Row"

139

Energy Flux We discuss various ways of describing energy flux and related quantities.  

E-Print Network (OSTI)

Chapter 6 Energy Flux We discuss various ways of describing energy flux and related quantities. 6.0.1 Energy Current Density The energy current density is given by the Poynting vector S = E ? H (6.1) where all quantities are real. The Poynting vector gives the instantaneous lo- cal energy current density

Palffy-Muhoray, Peter

140

Hedging Quantity Risks with Standard Power Options in a Competitive Wholesale Electricity  

E-Print Network (OSTI)

Hedging Quantity Risks with Standard Power Options in a Competitive Wholesale Electricity Market, GA, 30332-0205 USA March 3, 2005 Abstract This paper addresses quantity risk in the electricity of a load serving entity, which provides electricity service at a regulated price in electricity markets

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

Hedging Quantity Risks with Standard Power Options in a Competitive Wholesale Electricity Market  

E-Print Network (OSTI)

Hedging Quantity Risks with Standard Power Options in a Competitive Wholesale Electricity MarketScience (www.interscience.wiley.com). Abstract: This paper addresses quantity risk in the electricity market-serving entity, which provides electricity service at a regulated price in electricity markets with price

Oren, Shmuel S.

142

FINAL REPORT DETERMINATION OF THE PROCESSING RATE OF RPP WTP HLW SIMULANTS USING A DURAMELTER J 1000 VITRIFICATION SYSTEM VSL-00R2590-2 REV 0 8/21/00  

Science Conference Proceedings (OSTI)

This report provides data, analysis, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic University of America (VSL) to determine the melter processing rates that are achievable with RPP-WTP HLW simulants. The principal findings were presented earlier in a summary report (VSL-00R2S90-l) but the present report provides additional details. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. As a consequence of the limited amount of relevant information, there exists, for good reasons, a significant disparity between design-base specific glass production rates for the RPP-WTP LAW and HLW conceptual designs (1.0 MT/m{sup 2}/d and 0.4 MT/m{sup 2}/d, respectively); furthermore, small-scale melter tests with HLW simulants that were conducted during Part A indicated typical processing rates with bubbling of around 2.0 MT/m{sup 2}/d. This range translates into more than a factor of five variation in the resultant surface area of the HLW melter, which is clearly not without significant consequence. It is clear that an undersized melter is undesirable in that it will not be able to support the required waste processing rates. It is less obvious that there are potential disadvantages associated with an oversized melter, over and above the increased capital costs. A melt surface that is consistently underutilized will have poor cold cap coverage, which will result in increased volatilization from the melt (which is generally undesirable) and increased plenum temperatures due to increased thermal radiation from the melt surface (which mayor may not be desirable but the flexibility to choose may be lost). Increased volatilization is an issue both in terms of the increased challenge to the off-gas system as well as for the ability to effectively close the recycle loops for volatile species that must be immobilized in the glass product, most notably technetium and cesium. For these reasons, improved information is needed on the specific glass production rates of RPP-WTP HLW streams in DuraMelterJ systems over a range of operating conditions. Unlike the RPP-WTP LAW program, for which a pilot melter system to provide large-scale throughout information is already in operation, there is no comparable HLW activity; the results of the present study are therefore especially important. This information will reduce project risk by reducing the uncertainty associated with the amount of conservatism that mayor may not be associated with the baseline RPP-WTP HLW melter sizing decision. After the submission of the first Test Plan for this work, the RPP-WTP requested revisions to include tests to determine the processing rates that are achievable without bubbling, which was driven by the potential advantages of omitting bubblers from the HLW melter design in terms of reduced maintenance. A further objective of this effort became the determination of whether the basis of design processing rate could be achieved without bubbling. Ideally, processing rate tests would be conducted on a full-scale RPP-WTP melter system with actual HLW materials, but that is clearly unrealistic during Part B1. As a practical compromise the processing rate determinations were made with HL W simulants on a DuraMelter J system at as close to full scale as possible and the DM 1000 system at VSL was selected for that purpose. That system has a melt surface area of 1.2 m{sup 2}, which corresponds to about one-third scale based on the specific glass processing rate of 0.4 MT/m{sup 2}/d assumed in the RPP-WTP HLW conceptual design, but would correspon

KRUGER AA; MATLACK KS; KOT WK; PEREZ-CARDENAS F; PEGG IL

2011-12-29T23:59:59.000Z

143

Immobilization of Nuclear Wastes  

Science Conference Proceedings (OSTI)

Oct 20, 2010 ... Glassy and Glass Composite Nuclear Wasteforms: Michael Ojovan1; Bill Lee2; ... wastes which should be solidified for safe storage and disposal. ... has been vitrifying the Department of Energy's High Level Waste (HLW) at...

144

The Temporal Variability of Soil Moisture and Surface Hydrological Quantities in a Climate Model  

Science Conference Proceedings (OSTI)

The variance budget of land surface hydrological quantities is analyzed in the second Atmospheric Model Intercomparison Project (AMIP2) simulation made with the Canadian Centre for Climate Modelling and Analysis (CCCma) third-generation general ...

Vivek K. Arora; George J. Boer

2006-11-01T23:59:59.000Z

145

Some Correlations between the Large-Scale Meridional Eddy Momentum Transport and Zonal Mean Quantities  

Science Conference Proceedings (OSTI)

An empirical study has been made which compares the large-scale meridional eddy momentum transport with some selected zonal mean quantities by calculating correlations between them as a function of time lag and latitude. The basic dataset was the ...

Anne Leach

1984-01-01T23:59:59.000Z

146

Table 7.7 Quantity of Purchased Electricity, Natural Gas, and Steam, 2002  

U.S. Energy Information Administration (EIA) Indexed Site

7 Quantity of Purchased Electricity, Natural Gas, and Steam, 2002;" 7 Quantity of Purchased Electricity, Natural Gas, and Steam, 2002;" " Level: National and Regional Data;" " Row: NAICS Codes;" " Column: Supplier Sources of Purchased Electricity, Natural Gas, and Steam;" " Unit: Physical Units or Btu." ,,,"Electricity","Components",,"Natural Gas","Components",,"Steam","Components" " "," ",,,"Electricity",,,"Natural Gas",,,"Steam"," ",," " " "," ",,"Electricity","from Sources",,"Natural Gas","from Sources",,"Steam","from Sources"

147

Table A30. Quantity of Electricity Sold to Utility and Nonutility Purchasers  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Electricity Sold to Utility and Nonutility Purchasers" Quantity of Electricity Sold to Utility and Nonutility Purchasers" " by Census Region, Census Division, Industry Group, and Selected Industries, 1994" " (Estimates in Million Kilowatthours)" " "," "," "," "," ","RSE" "SIC"," "," ","Utility ","Nonutility","Row" "Code(a)","Industry Group and Industry","Total Sold","Purchaser(b)","Purchaser(c)","Factors" ,,"Total United States" ,"RSE Column Factors:",0.9,1.1,1 , 20,"Food and Kindred Products",1829," W "," W ",28

148

FCRD-USED-2010-000033, LLW Quantity and Inventory, FINAL R...  

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

Cycle Research and Development Cycle Research and Development Used Fuel Disposition Low Level Waste Disposition - Quantity and Inventory Prepared for U.S. Department of Energy Used Nuclear Fuel Robert H. Jones, SRS June 2011 Revision 2 FCRD-USED-2010-000033 FCRD-USED-2010-000033 Fuel Cycle Research and Development June 2011 Used Fuel Disposition Revision 2 Low Level Waste - Quantity and Inventory Page ii of x THIS PAGE INTENTIONALLY LEFT BLANK

149

HLW System Integrated Project Team  

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

l l W S Hi h l W S High Level Waste System High Level Waste System Integrated Project Team Integrated Project Team Integrated Project Team Integrated Project Team Steve Schneider Steve Schneider Office of Engineering and Technology High Level Waste Corporate Board March 5, 2009 This document is intended for planning and analysis purposes, assuming a continuing constrained budget environment. Every effort will be made to comply with all applicable environmental and legal obligations, while also assuring that essential functions necessary to protect human health, the environment and national security are maintained. 1 Introduction Introduction Introduction Introduction Challenges and Priorities High Level Waste Strategic Initiative Results High Level Waste System Integrated

150

Table A21. Quantity of Electricity Sold to Utility and Nonutility Purchasers  

U.S. Energy Information Administration (EIA) Indexed Site

1. Quantity of Electricity Sold to Utility and Nonutility Purchasers" 1. Quantity of Electricity Sold to Utility and Nonutility Purchasers" " by Census Region and Economic Characteristics of the Establishment, 1991" " (Estimates in Million Kilowatthours)" ,,,,"RSE" " "," ","Utility ","Nonutility","Row" "Economic Characteristics(a)","Total Sold","Purchaser(b)","Purchaser(c)","Factors" ,"Total United States",,, "RSE Column Factors:",1,1.1,1 "Value of Shipments and Receipts" "(million dollars)" " Under 20",188,122,66,35.6 " 20-49",2311,1901,410,39.5 " 50-99",2951,2721,230,9.6 " 100-249",6674,5699,974,7.1

151

Table A31. Quantity of Electricity Sold to Utility and Nonutility Purchasers  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Electricity Sold to Utility and Nonutility Purchasers by Census Region," Quantity of Electricity Sold to Utility and Nonutility Purchasers by Census Region," " Census Division, and Economic Characteristics of the Establishment, 1994" " (Estimates in Million Kilowatthours)" ,,,,"RSE" " "," ","Utility ","Nonutility","Row" "Economic Characteristics(a)","Total Sold","Purchaser(b)","Purchaser(c)","Factors" ,"Total United States",,, "RSE Column Factors:",0.9,1.1,1 "Value of Shipments and Receipts" "(million dollars)" " Under 20",222,164," Q ",23.3 " 20-49",1131,937,194,17.2

152

"Table A32. Total Quantity of Purchased Energy Sources by Census Region,"  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Purchased Energy Sources by Census Region," Quantity of Purchased Energy Sources by Census Region," " Census Division, Industry Group, and Selected Industries, 1994" " (Estimates in Btu or Physical Units)" ,,,,,,"Natural",,,"Coke" " "," ","Total","Electricity","Residual","Distillate","Gas(c)"," ","Coal","and Breeze"," ","RSE" "SIC"," ","(trillion","(million","Fuel Oil","Fuel Oil(b)","(billion","LPG","(1000","(1000","Other(d)","Row" "Code(a)","Industry Group and Industry","Btu)","kWh)","(1000 bbl)","(1000 bbl)","cu ft)","(1000 bbl)","short tons)","short tons)","(trillion Btu)","Factors"

153

Table A18. Quantity of Electricity Sold to Utility and Nonutility Purchasers  

U.S. Energy Information Administration (EIA) Indexed Site

8. Quantity of Electricity Sold to Utility and Nonutility Purchasers" 8. Quantity of Electricity Sold to Utility and Nonutility Purchasers" " by Census Region, Industry Group, and Selected Industries, 1991" " (Estimates in Million Kilowatthours)" " "," "," "," "," ","RSE" "SIC"," "," ","Utility ","Nonutility","Row" "Code(a)","Industry Groups and Industry","Total Sold","Purchaser(b)","Purchaser(c)","Factors" ,,"Total United States" ,"RSE Column Factors:",0.9,1,1 , 20,"Food and Kindred Products",988,940,48,16.2 2011," Meat Packing Plants",0,0,0,"NF"

154

"Table A22. Total Quantity of Purchased Energy Sources by Census Region,"  

U.S. Energy Information Administration (EIA) Indexed Site

2. Total Quantity of Purchased Energy Sources by Census Region," 2. Total Quantity of Purchased Energy Sources by Census Region," " Industry Group, and Selected Industries, 1991" " (Estimates in Btu or Physical Units)" ,,,,,,"Natural",,,"Coke" " "," ","Total","Electricity","Residual","Distillate","Gas(c)"," ","Coal","and Breeze"," ","RSE" "SIC"," ","(trillion","(million","Fuel Oil","Fuel Oil(b)","(billion","LPG","(1000","(1000","Other(d)","Row" "Code(a)","Industry Groups and Industry","Btu)","kWh)","(1000 bbls)","(1000 bbls)","cu ft)","(1000 bbls)","short tons)","short tons)","(trillion Btu)","Factors"

155

Proposal for a quantity based data model in the Virtual Observatory  

E-Print Network (OSTI)

We propose the beginnings of a data model for the Virtual Observatory (VO) built up from simple ``quantity'' objects. In this paper we present how an object-oriented, domain (or namespace)-scoped simple quantity may be used to describe astronomical data. Our model is designed around the requirements that it be searchable and serve as a transport mechanism for all types of VO data and meta-data. In this paper we describe this model in terms of an OWL ontology and UML diagrams. An XML schema is available online.

Brian Thomas; Edward Shaya

2003-12-23T23:59:59.000Z

156

Characteristics Data Base: Programmer's guide to the LWR Quantities Data Base  

SciTech Connect

The LWR Quantities Data Base is a menu-driven PC data base developed as part of OCRWM's waste, technical data base on the characteristics of potential repository wastes, which also includes non-LWR spent fuel, high-level and other materials. This programmer's guide completes the documentation for the LWR Quantities Data Base, the user's guide having been published previously. The PC data base itself may be requested from the Oak Ridge National Laboratory, using the order form provided in Volume 1 of publication DOE/RW-0184.

Jones, K.E. (DataPhile, Inc., Knoxville, TN (USA)); Moore, R.S. (Automated Sciences Group, Inc., Oak Ridge, TN (USA))

1990-08-01T23:59:59.000Z

157

The Monitoring System for Electric Quantity Consumed in Extruder Based on WB Electrical Transducer  

Science Conference Proceedings (OSTI)

A new system was discussed, which can be used to measure the performance parameters of extruders driven by asynchronous motor. By applying the WB electrical transducer and the electrical power method, the problem such as low-precision measurement is ... Keywords: electric quantity measurement, electrical transducer, extruder, asynchronous motor

Zheng Shuang; Liu Fugang

2010-06-01T23:59:59.000Z

158

A comparison of cloud microphysical quantities with forecasts from cloud prediction models  

SciTech Connect

Numerical weather prediction models (ECMWF, NCEP) are evaluated using ARM observational data collected at the Southern Great Plains (SGP) site. Cloud forecasts generated by the models are compared with cloud microphysical quantities, retrieved using a variety of parameterizations. Information gained from this comparison will be utilized during the FASTER project, as models are evaluated for their ability to reproduce fast physical processes detected in the observations. Here the model performance is quantified against the observations through a statistical analysis. Observations from remote sensing instruments (radar, lidar, radiometer and radiosonde) are used to derive the cloud microphysical quantities: ice water content, liquid water content, ice effective radius and liquid effective radius. Unfortunately, discrepancies in the derived quantities arise when different retrieval schemes are applied to the observations. The uncertainty inherent in retrieving the microphysical quantities using various retrievals is estimated from the range of output microphysical values. ARM microphysical retrieval schemes (Microbase, Mace) are examined along with the CloudNet retrieval processing of data from the ARM sites for this purpose. Through the interfacing of CloudNet and ARM processing schemes an ARMNET product is produced and employed as accepted observations in the assessment of cloud model predictions.

Dunn, M.; Jensen, M.; Hogan, R.; OConnor, E.; Huang, D.

2010-03-15T23:59:59.000Z

159

FINAL REPORT REGULATORY OFF GAS EMISSIONS TESTING ON THE DM1200 MELTER SYSTEM USING HLW AND LAW SIMULANTS VSL-05R5830-1 REV 0 10/31/05  

Science Conference Proceedings (OSTI)

The operational requirements for the River Protection Project - Waste Treatment Plant (RPP-WTP) Low Activity Waste (LAW) and High Level Waste (HLW) melter systems, together with the feed constituents, impose a number of challenges to the off-gas treatment system. The system must be robust from the standpoints of operational reliability and minimization of maintenance. The system must effectively control and remove a wide range of solid particulate matter, acid mists and gases, and organic constituents (including those arising from products of incomplete combustion of sugar and organics in the feed) to concentration levels below those imposed by regulatory requirements. The baseline design for the RPP-WTP LAW primary off-gas system includes a submerged bed scrubber (SBS), a wet electrostatic precipitator (WESP), and a high efficiency particulate air (HEPA) filter. The secondary off-gas system includes a sulfur-impregnated activated carbon bed (AC-S), a thermal catalytic oxidizer (TCO), a single-stage selective catalytic reduction NOx treatment system (SCR), and a packed-bed caustic scrubber (PBS). The baseline design for the RPP-WTP HLW primary off-gas system includes an SBS, a WESP, a high efficiency mist eliminator (HEME), and a HEPA filter. The HLW secondary off-gas system includes a sulfur-impregnated activated carbon bed, a silver mordenite bed, a TCO, and a single-stage SCR. The one-third scale HLW DM1200 Pilot Melter installed at the Vitreous State Laboratory (VSL) was equipped with a prototypical off-gas train to meet the needs for testing and confirmation of the performance of the baseline off-gas system design. Various modifications have been made to the DM1200 system as the details of the WTP design have evolved, including the installation of a silver mordenite column and an AC-S column for testing on a slipstream of the off-gas flow; the installation of a full-flow AC-S bed for the present tests was completed prior to initiation of testing. The DM1200 system was reconfigured to enable testing of the baseline HLW or LAW off-gas trains to perform off-gas emissions testing with both LAW and HLW simulants in the present work. During 2002 and 2003, many of these off-gas components were tested individually and in an integrated manner with the DM1200 Pilot Melter. Data from these tests are being used to support engineering design confirmation and to provide data to support air permitting activities. In fiscal year 2004, the WTP Project was directed by the Office of River Protection (ORP) to comply with Environmental Protection Agency (EPA) Maximum Achievable Control Technology (MACT) requirements for organics. This requires that the combined melter and off-gas system have destruction and removal efficiency (DRE) of >99.99% for principal organic dangerous constituents (PODCs). In order to provide confidence that the melter and off-gas system are able to achieve the required DRE, testing has been directed with both LAW and HLW feeds. The tests included both 'normal' and 'challenge' WTP melter conditions in order to obtain data for the potential range of operating conditions for the WTP melters and off-gas components. The WTP Project, Washington State Department of Ecology, and ORP have agreed that naphthalene will be used for testing to represent semi-volatile organics and allyl alcohol will be used to represent volatile organics. Testing was also performed to determine emissions of halides, metals, products of incomplete combustion (PICs), dioxins, furans, coplanar PCBs, total hydrocarbons, and COX and NOX, as well as the particle size distribution (PSD) of particulate matter discharged at the end of the off-gas train. A description of the melter test requirements and analytical methods used is provided in the Test Plan for this work. Test Exceptions were subsequently issued which changed the TCO catalyst, added total organic emissions (TOE) to exhaust sampling schedule, and allowing modification of the test conditions in response to attainable plenum temperatures as well as temperature increases in the sulfur impr

KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D'ANGELO NA; BRANDYS M; KOT WK; PEGG IL

2011-12-29T23:59:59.000Z

160

Final Report - Glass Formulation Development and Testing for DWPF High AI2O3 HLW Sludges, VSL-10R1670-1, Rev. 0, dated 12/20/10  

Science Conference Proceedings (OSTI)

The principal objective of the work described in this Final Report is to develop and identify glass frit compositions for a specified DWPF high-aluminum based sludge waste stream that maximizes waste loading while maintaining high production rate for the waste composition provided by ORP/SRS. This was accomplished through a combination of crucible-scale, vertical gradient furnace, and confirmation tests on the DM100 melter system. The DM100-BL unit was selected for these tests. The DM100-BL was used for previous tests on HLW glass compositions that were used to support subsequent tests on the HLW Pilot Melter. It was also used to process compositions with waste loadings limited by aluminum, bismuth, and chromium, to investigate the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition, to process glass formulations at compositional and property extremes, and to investigate crystal settling on a composition that exhibited one percent crystals at 963{degrees}C (i.e., close to the WTP limit). The same melter was selected for the present tests in order to maintain comparisons between the previously collected data. The tests provide information on melter processing characteristics and off-gas data, including formation of secondary phases and partitioning. Specific objectives for the melter tests are as follows: Determine maximum glass production rates without bubbling for a simulated SRS Sludge Batch 19 (SB19). Demonstrate a feed rate equivalent to 1125 kg/m{sup 2}/day glass production using melt pool bubbling. Process a high waste loading glass composition with the simulated SRS SB19 waste and measure the quality of the glass product. Determine the effect of argon as a bubbling gas on waste processing and the glass product including feed processing rate, glass redox, melter emissions, etc.. Determine differences in feed processing and glass characteristics for SRS SB19 waste simulated by the co-precipitated and direct-hydroxide methods. The above tests were proposed based on previous tests for WTP in which there were few differences in the melter processing characteristics, such as processing rate and melter emissions, between precipitated and direct hydroxide simulants, even though there were differences in rheological properties. To the extent this similarity is found also for simulants for SRS HLW, the direct hydroxide methods may offer the potential for faster, simpler, and cheaper simulant production. There was no plan to match the yield stress and particle size of the direct hydroxide simulant to that of the precipitated simulant because that would have increased the preparation cost and complexity and defeated the purpose of the tests. These objectives were addressed by first developing a series of glass frits and then conducting a crucible scale study to determine the waste loading achievable for the waste composition and to select the preferred frit. Waste loadings were increased until the limits of a glass property were exceeded experimentally. Glass properties for evaluation included: viscosity, electrical conductivity, crystallinity (including liquidus temperature and nepheline formation after canister centerline cooling (CCC) heat-treatment), gross glass phase separation, and the 7- day Product Consistency Test (PCT, ASTM-1285) response. Glass property limits were based upon the constraints used for DWPF process control.

Kruger, Albert A.; Pegg, I. L.; Kot, W. K.; Gan, H.; Matlack, K. S.

2013-11-13T23:59:59.000Z

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

SAFETY ANALYSIS REPORT FOR PACKAGING, MODEL 9977, ADDENDUM 3, JUSTIFICATION FOR SMALL GRAM QUANTITY CONTENTS  

SciTech Connect

This Addendum establishes a new family of content envelopes consisting of small quantities of radioactive materials. These content envelopes and specific packing configurations are shown to be subcritical. However, the dose rates of some payloads must be measured and shown to comply with applicable radiation limits. Authorization for shipment of the content envelop requires acceptance of this Addendum by the DOE-HQ certifying official as a supplement to the 9977 SARP Revision 2 and DOE-HQ?s subsequent revision of the CoC Revision 10 (which is based on SARP Addendum 2 and SARP Addendum 4) to authorize the additional content envelope. The Small Gram Quantity Content Envelopes and packing configurations will be incorporated in the next revision of the 9977 SARP.

Abramczyk, G.

2011-10-31T23:59:59.000Z

162

Table 7.7 Quantity of Purchased Electricity, Natural Gas, and Steam, 2010;  

U.S. Energy Information Administration (EIA) Indexed Site

7 Quantity of Purchased Electricity, Natural Gas, and Steam, 2010; 7 Quantity of Purchased Electricity, Natural Gas, and Steam, 2010; Level: National and Regional Data; Row: NAICS Codes; Column: Supplier Sources of Purchased Electricity, Natural Gas, and Steam; Unit: Physical Units or Btu. Electricity Components Natural Gas Components Steam Components Electricity Natural Gas Steam Electricity from Sources Natural Gas from Sources Steam from Sources Electricity from Local Other than Natural Gas from Local Other than Steam from Local Other than NAICS Total Utility(b) Local Utility(c) Total Utility(b) Local Utility(c) Total Utility(b) Local Utility(c) Code(a) Subsector and Industry (million kWh) (million kWh) (million kWh) (billion cu ft) (billion cu ft)

163

Shipment of Small Quantities of Unspecified Radioactive Material in Chalfant Packagings  

SciTech Connect

In the post 6M era, radioactive materials package users are faced with the disciplined operations associated with use of Certified Type B packagings. Many DOE, commercial and academic programs have a requirement to ship and/or store small masses of poorly characterized or unspecified radioactive material. For quantities which are small enough to be fissile exempt and have low radiation levels, the materials could be transported in a package which provides the required containment level. Because their Chalfant type containment vessels meet the highest standard of containment (helium leak-tight), the 9975, 9977, and 9978 are capable of transporting any of these contents. The issues associated with certification of a high-integrity, general purpose package for shipping small quantities of unspecified radioactive material are discussed and certification of the packages for this mission is recommended.

Smith, Allen; Abramczyk, Glenn; Nathan, Steven; Bellamy, Steve

2009-06-12T23:59:59.000Z

164

An indirect sensing technique for diesel fuel quantity control. Progress report, April 1--June 30, 1998  

DOE Green Energy (OSTI)

This reports on a project to develop an indirect sensing technique for diesel fuel quantity control. Development has continued on a vehicle-installed prototype for EPA certification and demonstration. Focus of development is on the use of this technology for retrofitting existing diesel vehicles to reduce emissions rather than exclusively upon deployment in the OEM market. Technical obstacles that have been encountered and their solutions and remaining project tasks are described.

MacCarley, C.A.

1998-08-31T23:59:59.000Z

165

Unrestricted EM Nation-Wide Indefinite Delivery/Indefinite Quantity Contracts  

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

UNRESTRICTED EM NATION-WIDE UNRESTRICTED EM NATION-WIDE INDEFINITE DELIVERY/INDEFINITE QUANTITY CONTRACTS Contractor Name Principal Subcontracts AECOM Technical Services, Inc. Bartlett Services, Inc., Cavanagh Services Group, Inc. Clauss Construction EnergX, LLC NuVision Engineering Bechtel National, Inc. Eberline Services, Inc. North Wind, Inc. Philotechnics, Ltd. TC Program Solutions, LLC CDM, JV Navarro Research and Engineering, Inc. Newport News Nuclear, Inc. MSE Technology Applications, Inc. CH2M Hill Constructors, Inc. Babcock & Wilcox Technical Services Group EnergySolutions Federal Solutions, Inc. DEMCO, Inc. Terranear PMC, LLC Tetra Tech EC, Inc. Fluor Federal Services, Inc.

166

Set-Aside EM Nation-Wide Indefinite Delivery/Indefinite Quantity Contracts  

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

SET-ASIDE EM NATION-WIDE SET-ASIDE EM NATION-WIDE INDEFINITE DELIVERY/INDEFINITE QUANTITY CONTRACTS Contractor Name Team Members/Principal Subcontractors Clauss Construction AECOM, Inc. Cavanagh Services Group, Inc. EnergX, LLC Dynamic Management Solutions, LLC (DMS) CA - LLC Member Restoration Services, Inc, LLC Member Wastren Advantage, Inc. - LLC Member SAIC Bartlett Services, Inc. Siempelkamp Nuclear Services Gonzales-Stoller Remediation Services, LLC JG Management Systems, Inc. - LLC Member (Protégé) The S.M. Stoller Corporation - LLC Member (Mentor) AET Environmental, Inc. ALS Laboratory Group AquaTierra Associates, Inc. DBA Weiss Assoc. ARCADIS U.S., Inc. AREVA Federal Services, LLC

167

Method for determining trace quantities of chloride in polymeric materials using ion selective electrodes: Final report  

Science Conference Proceedings (OSTI)

A method for determining trace quantities of chloride in polymeric materials has been developed. Ion-selective electrodes and the standard addition method were used in all the analyses. The ion-selective electrode method was compared with neutron activation, ion chromatography and chloridometer titration. The ion-selective electrode technique results for chloride were similar to those of neutron activation, which is the acknowledged referee method. This ion-selective electrode method showed the highest standard recovery when compared with the ion chromatography and chloridometer titration methods.

Salary, J.

1987-02-01T23:59:59.000Z

168

FINAL REPORT MELTER TESTS WITH AZ-101 HLW SIMULANT USING A DURAMELTER 100 VITRIFICATION SYSTEM VSL-01R10N0-1 REV 1 2/25/02  

Science Conference Proceedings (OSTI)

This report provides data, analyses, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic of America (VSL) to determine the processing rates that are achievable with AZ-101 HLW simulants and corresponding melter feeds on a DuraMelter 100 (DM100) vitrification system. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. Tests conducted during Part B1 (VSL-00R2590-2) on the DM1000 vitrification system installed at the Vitreous State Laboratory of The Catholic University of America showed that, without the use of bubblers, glass production rates with AZ-101 and C-106/AY-102 simulants were significantly lower than the Project design basis rate of 0.4 MT/m{sup 2}/d. Conversely, three-fold increases over the design basis rate were demonstrated with the use of bubblers. Furthermore, an un-bubbled control test using a replica of the melter feed used in cold commissioning tests at West Valley reproduced the rates that were observed with that feed on the WVDP production melter. More recent tests conducted on the DM1200 system, which more closely represents the present RPP-WTP design, are in general agreement with these earlier results. Screening tests conducted on the DM10 system have provided good indications of the larger-scale processing rates with bubblers (for both HL W and LAW feeds) but significantly overestimated the DM1000 un-bubbled rate observed for C-106/AY-102 melter feeds. This behavior is believed to be a consequence of the role of heat transfer in rate attainment and the much greater role of wall effects in heat transfer when the melt pool is not agitated. The DM100 melter used for the present tests has a surface area of 0.108 m{sup 2}, which is approximately 5 times larger than that of the DM10 (0.021 m{sup 2}) and approximately 11 times smaller than that of the DM1000 (1.2 m{sup 2}) (the DM1000 has since been replaced by a pilot-scale prototypical HLW melter, designated the DM1200, which has the same surface area as the DM1000). Testing on smaller melters is the most economical method for obtaining data over a wide range of operating conditions (particularly at extremes) and for guiding the more expensive tests that are performed at pilot-scale. Thus, one objective of these tests was to determine whether the DM100 melters are sufficiently large to reproduce the un-bubbled melt rates observed at the DM1000 scale, or to determine the extent of any off-set. DM100-scale tests can then be used to screen feed chemistry variations that may serve to increase the un-bubbled production rates prior to confirmation at pilot scale. Finally, extensive characterization data obtained on simulated HLW melter feeds formed from various glass forming additives indicated that there may be advantages in terms of feed rheology and stability to the replacement of some of the hydroxides by carbonates. A further objective of the present tests was therefore to identify any deleterious processing effects of such a change before adopting the carbonate feed as the baseline. Data from the WVDP melter using acidified (nitrated) feeds, and without bubbling, showed productions rates that are higher than those observed with the alkaline RPP feeds at the VSL. Therefore, the effect of feed acidification on production rate also was investigated. This work was performed under Test Specification, 'TSP-W375-00-00019, Rev 0, 'HLW-DM10 and DM100 Melter Tests' dated November 13, 2000 and the corresponding Test Plan. It should be noted, however, that the RPP-WTP Project directed a series of changes to the Test Plan as the result

KRUGER AA; MATLACK KS; KOT WK; PEGG IL

2011-12-29T23:59:59.000Z

169

NETL: News Release - New Projects to Study Ways to Recover Vast Quantities  

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

March 12, 2002 March 12, 2002 New Projects to Study Ways to Recover Vast Quantities of "Left Behind" Oil TULSA, OK - Nearly two out of every three barrels of oil discovered in the United States remain trapped underground after conventional recovery operations. This staggering amount of remaining oil - approximately 200 billion barrels - can be one of America's best hopes for greater energy security if new technologies can be developed to recover it. Often, however, the "left behind" oil is in regions of the reservoir that are difficult to access and the oil is held tightly in place within tiny rock pores by capillary pressures that resist many traditional oil production practices. Now, as part of its program to develop ways to free this unrecovered oil, the Department of Energy's Fossil Energy research program is adding three new projects to be carried out by three of the Nation's top petroleum engineering universities:

170

Department of Energy, Indefinite Delivery Indefinite Quantity, Multiple Award, Energy Savings Performance  

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

DE-AM36-09GO290XX / Mod 6 DE-AM36-09GO290XX / Mod 6 Department of Energy, Indefinite Delivery Indefinite Quantity, Multiple Award, Energy Savings Performance Contract Awarded by the Department of Energy, Golden Field Office to 16 Energy Service Companies (ESCOs) on December 17, 2008 What follows is a generic version of the contract. All 16 contracts are identical with the exception of the SF33, Solicitation, Offer and Award, and Attachment J-13, Subcontracting Plan, which are specific to each ESCO. This is a conformed version of the contract as of November 2012, through modification 6, excluding contractor specific modifications for novations and name changes." 1 NEGOTIATED (RFP) November 2012 DE-AM36-09GO290XX / Mod 6 SOLICITATION, OFFER, AND AWARD 1. THIS CONTRACT IS A RATED ORDER

171

Department of Energy, Indefinite Delivery Indefinite Quantity, Multiple Award, Energy Savings Performance  

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

2 DE-AM36-09GO290XX / Mod 6 2 DE-AM36-09GO290XX / Mod 6 Department of Energy, Indefinite Delivery Indefinite Quantity, Multiple Award, Energy Savings Performance Contract Awarded by the Department of Energy, Golden Field Office to 16 Energy Service Companies (ESCOs) on December 17, 2008 What follows is a generic version of the contract. All 16 contracts are identical with the exception of the SF33, Solicitation, Offer and Award, and Attachment J-13, Subcontracting Plan, which are specific to each ESCO. This is a conformed version of the contract as of November 2012, through modification 6, excluding contractor specific modifications for novations and name changes." 1 NEGOTIATED (RFP) November 2012 DE-AM36-09GO290XX / Mod 6 SOLICITATION, OFFER, AND AWARD 1. THIS CONTRACT IS A RATED ORDER

172

Methane storage in multi-walled carbon nanotubes at the quantity of 80 g  

SciTech Connect

Methane storage in multi-walled carbon nanotubes (MWNTs) is studied at ambient temperature and pressures of 0-10.5 MPa, with a quantity of 80 g samples that were synthesized by nano-agglomerate fluidized-bed reactors (NAFBR). The volume of methane released by MWNTs was measured by volumetric method. We study the effects of purification and the pretreatments on methane storage. Results show that mixed acid treatment, alkali treatment, and mechanical shearing can obviously enhance gas uptake while high-temperature treatment can only slightly reduce it. For properly pretreated samples, an optimal 11.7% of mass storage capacity was achieved at room temperature and the pressure of 10.5 MPa, indicating that CNTs is a potential material for methane uptake.

Wu Yulong [Department of Chemical Engineering, Tsinghua University, Beijing 100084 (China); Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Wei Fei [Department of Chemical Engineering, Tsinghua University, Beijing 100084 (China)], E-mail: wf-dce@tsinghua.edu.cn; Luo Guohua; Ning Guoqing [Department of Chemical Engineering, Tsinghua University, Beijing 100084 (China); Yang Mingde [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

2008-06-03T23:59:59.000Z

173

Effects of space velocity on quality and quantity of gas from oil shale retorting  

DOE Green Energy (OSTI)

The effects of gas injection rate, water injection rate, and retort pressure on offgas quality and quantity were studied using a small laboratory retort. There are interactions between these variables and oxygen injection and oil shale grade which affect total energy recovery, oil recovery, and energy content of the offgas. Prediction equations were developed describing the effects of these interactions. The study shows that with a low gas injection rate of 0.5 scfm/ft/sup 2/ at 100 psig with a water injection rate of 0.0201 lb/ft/sup 2/-min 50 percent of the potential oil recovery can be obtained together with 250 Btu/ft/sup 3/ gas.

Jacobson, I.A. Jr.; Burwell, E.L.

1976-04-01T23:59:59.000Z

174

Joint production and economic retention quantity decisions in capacitated production systems serving multiple market segments  

E-Print Network (OSTI)

In this research, we consider production/inventory management decisions of a rmthat sells its product in two market segments during a nite planning horizon. In thebeginning of each period, the rm makes a decision on how much to produce basedon the production capacity and the current on-hand inventory available. After theproduction is made at the beginning of the period, the rm rst satises the stochasticdemand from customers in its primary market. Any primary market demand thatcannot be satised is lost. After satisfying the demand from the primary market, ifthere is still inventory on hand, all or part of the remaining products can be sold ina secondary market with ample demand at a lower price. Hence, the second decisionthat the rm makes in each period is how much to sell in the secondary market, orequivalently, how much inventory to carry to the next period.The objective is to maximize the expected net revenue during a nite planninghorizon by determining the optimal production quantity in each period, and theoptimal inventory amount to carry to the next period after the sales in primary andsecondary markets. We term the optimal inventory amount to be carried to the nextperiod as \\economic retention quantity". We model this problem as a nite horizonstochastic dynamic program. Our focus is to characterize the structure of the optimalpolicy and to analyze the system under dierent parameter settings. Conditioning on given parameter set, we establish lower and upper bounds on the optimal policyparameters. Furthermore, we provide computational tools to determine the optimalpolicy parameters. Results of the numerical analysis are used to provide furtherinsights into the problem from a managerial perspective.

Katariya, Abhilasha Prakash

2008-08-01T23:59:59.000Z

175

PCP METHODOLOGY FOR DETERMINING DOSE RATES FOR SMALL GRAM QUANTITIES IN SHIPPING PACKAGINGS  

Science Conference Proceedings (OSTI)

The Small Gram Quantity (SGQ) concept is based on the understanding that small amounts of hazardous materials, in this case radioactive materials, are significantly less hazardous than large amounts of the same materials. This study describes a methodology designed to estimate an SGQ for several neutron and gamma emitting isotopes that can be shipped in a package compliant with 10 CFR Part 71 external radiation level limits regulations. These regulations require packaging for the shipment of radioactive materials perform, under both normal and accident conditions, the essential functions of material containment, subcriticality, and maintain external radiation levels within regulatory limits. 10 CFR 71.33(b)(1)(2)&(3) state radioactive and fissile materials must be identified and their maximum quantity, chemical and physical forms be included in an application. Furthermore, the U.S. Federal Regulations require application contain an evaluation demonstrating the package (i.e., the packaging and its contents) satisfies the external radiation standards for all packages (10 CFR 71.31(2), 71.35(a), & 71.47). By placing the contents in a He leak-tight containment vessel, and limiting the mass to ensure subcriticality, the first two essential functions are readily met. Some isotopes emit sufficiently strong photon radiation that small amounts of material can yield a large external dose rate. Quantifying of the dose rate for a proposed content is a challenging issue for the SGQ approach. It is essential to quantify external radiation levels from several common gamma and neutron sources that can be safely placed in a specific packaging, to ensure compliance with federal regulations. The Packaging Certification Program (PCP) Methodology for Determining Dose Rate for Small Gram Quantities in Shipping Packagings described in this report provides bounding mass limits for a set of proposed SGQ isotopes. Methodology calculations were performed to estimate external radiation levels for the 9977 shipping package using the MCNP radiation transport code to develop a set of response multipliers (Green's functions) for 'dose per particle' for each neutron and photon spectral group. The source spectrum for each isotope generated using the ORIGEN-S and RASTA computer codes was folded with the response multipliers to generate the dose rate per gram of each isotope in the 9977 shipping package and its associated shielded containers. The maximum amount of a single isotope that could be shipped within the regulatory limits contained in 10 CFR 71.47 for dose rate at the surface of the package is determined. If a package contains a mixture of isotopes, the acceptability for shipment can be determined by a sum of fractions approach. Furthermore, the results of this analysis can be easily extended to additional radioisotopes by simply evaluating the neutron and/or photon spectra of those isotopes and folding the spectral data with the Green's functions provided.

Nathan, S.

2011-08-23T23:59:59.000Z

176

PACKAGING CERTIFICATION PROGRAM METHODOLOGY FOR DETERMINING DOSE RATES FOR SMALL GRAM QUANTITIES IN SHIPPING PACKAGINGS  

Science Conference Proceedings (OSTI)

The Small Gram Quantity (SGQ) concept is based on the understanding that small amounts of hazardous materials, in this case radioactive materials (RAM), are significantly less hazardous than large amounts of the same materials. This paper describes a methodology designed to estimate an SGQ for several neutron and gamma emitting isotopes that can be shipped in a package compliant with 10 CFR Part 71 external radiation level limits regulations. These regulations require packaging for the shipment of radioactive materials, under both normal and accident conditions, to perform the essential functions of material containment, subcriticality, and maintain external radiation levels within the specified limits. By placing the contents in a helium leak-tight containment vessel, and limiting the mass to ensure subcriticality, the first two essential functions are readily met. Some isotopes emit sufficiently strong photon radiation that small amounts of material can yield a large dose rate outside the package. Quantifying the dose rate for a proposed content is a challenging issue for the SGQ approach. It is essential to quantify external radiation levels from several common gamma and neutron sources that can be safely placed in a specific packaging, to ensure compliance with federal regulations. The Packaging Certification Program (PCP) Methodology for Determining Dose Rate for Small Gram Quantities in Shipping Packagings provides bounding shielding calculations that define mass limits compliant with 10 CFR 71.47 for a set of proposed SGQ isotopes. The approach is based on energy superposition with dose response calculated for a set of spectral groups for a baseline physical packaging configuration. The methodology includes using the MCNP radiation transport code to evaluate a family of neutron and photon spectral groups using the 9977 shipping package and its associated shielded containers as the base case. This results in a set of multipliers for 'dose per particle' for each spectral group. For a given isotope, the source spectrum is folded with the response for each group. The summed contribution from all isotopes determines the total dose from the RAM in the container.

Nathan, S.; Loftin, B.; Abramczyk, G.; Bellamy, S.

2012-05-09T23:59:59.000Z

177

ITEM NO. SUPPLIES/SERVICES QUANTITY UNIT UNIT PRICE AMOUNT NAME OF OFFEROR OR CONTRACTOR  

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

ITEM NO. ITEM NO. SUPPLIES/SERVICES QUANTITY UNIT UNIT PRICE AMOUNT NAME OF OFFEROR OR CONTRACTOR 2 2 CONTINUATION SHEET REFERENCE NO. OF DOCUMENT BEING CONTINUED PAGE OF OAK RIDGE ASSOCIATED UNIVERSITIES, INC. (A) (B) (C) (D) (E) (F) DE-AC05-06OR23100/0456 Payment: OR for Oak Ridge/OSTI U.S. Department of Energy Oak Ridge Office Oak Ridge Financial Service Center P.O. Box 6017 Oak Ridge TN 37831 Period of Performance: 01/01/2006 to 12/31/2015 NSN 7540-01-152-8067 OPTIONAL FORM 336 (4-86) Sponsored by GSA FAR (48 CFR) 53.110 ___________ (x) x DE-AC05-06OR23100 copies of the amendment; (b) By acknowledging receipt of this amendment on each copy of the offer submitted; or (c) By separate letter or telegram which includes a reference to the solicitation and amendment numbers. FAILURE OF YOUR ACKNOWLEDGEMENT TO BE RECEIVED AT

178

Relation Between the Adsorbed Quantity and the Immersion Enthalpy in Catechol Aqueous Solutions on Activated Carbons  

E-Print Network (OSTI)

Abstract: An activated carbon, Carbochem TM PS230, was modified by chemical and thermal treatment in flow of H2, in order to evaluate the influence of the activated carbon chemical characteristics in the adsorption of the catechol. The catechol adsorption in aqueous solution was studied along with the effect of the pH solution in the adsorption process of modified activated carbons and the variation of immersion enthalpy of activated carbons in the aqueous solutions of catechol. The interaction solid-solution is characterized by adsorption isotherms analysis, at 298 K and pH 7, 9 and 11 in order to evaluate the adsorption value above and below that of the catechol pKa. The adsorption capacity of carbons increases when the solution pH decreases. The retained amount increases slightly in the reduced carbon to maximum adsorption pH and diminishes in the oxidized carbon. Similar conclusions are obtained from the immersion enthalpies, whose values increase with the solute quantity retained. In granular activated carbon (CAG), the immersion enthalpies obtained are between 21.5 and 45.7 Jg ?1 for catechol aqueous solutions in a range of 20 at 1500 mgL ?1.

Juan Carlos Moreno-pirajn; Diego Blanco; Liliana Giraldo

2011-01-01T23:59:59.000Z

179

Density equation of bio-coal briquettes and quantity of maize cob in Phitsanulok, Thailand  

SciTech Connect

One of the most important crops in Phitsanulok, a province in Northern Thailand, is maize. BaseD on the calculation, the quantity of maize cob produced in this region was approximately 220 kton year{sup -1}. The net heating value of maize cob was found to be 14.2 MJ kg{sup -1}. Therefore, the total energy over 874 TJ year-1 can be obtained from this agricultural waste. In the experiments, maize cob was utilized as the major ingredient for producing biomass-coal briquettes. The maize cob was treated with sodium hydroxide solution before mixing with coal fine. The ratios of coal:maize were 1:2 and 1:3, respectively. The range of briquetting pressures was from 4-8 MPa. The result showed that the density was strongly affected by both parameters. Finally, the relationship between biomass ratio, briquetting pressures and briquette density was developed and validated by using regression technique. 13 refs., 2 figs.

Patomsok Wilaipon [Naresuan University, Phitsanulok (Thailand). Department of Mechanical Engineering

2008-07-01T23:59:59.000Z

180

The composition of a quad of buildings sector energy: Physical, economic, and environmental quantities  

SciTech Connect

In an analysis conducted for the US Department of Energy Office of Building Technologies (OBT), the Pacific Northwest Laboratory examined the fuel type composition of energy consumed in the US buildings sector. Numerical estimates were developed for the physical quantities of fuel consumed, as well as of the fossil fuel emissions (carbon dioxide, sulfur dioxide, nitrogen oxides) and nuclear spent fuel byproducts associated with that consumption. Electric generating requirements and the economic values associated with energy consumption also were quantified. These variables were quantified for a generic quad (1 quadrillion Btu) of primary energy for the years 1987 and 2010, to illustrate the impacts of a fuel-neutral reduction in buildings sector energy use, and for specific fuel types, to enable meaningful comparisons of benefits achievable through various OBT research projects or technology developments. Two examples are provided to illustrate how these conversion factors may be used to quantify the impacts of energy savings potentially achievable through OBT building energy conservation efforts. 18 refs., 6 figs., 16 tabs.

Secrest, T.J.; Nicholls, A.K.

1990-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "vitrified hlw quantities" 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.


181

A wireless interrogation system exploiting narrowband acoustic resonator for remote physical quantity measurement  

SciTech Connect

Monitoring physical quantities using acoustic wave devices can be advantageously achieved using the wave characteristic dependence to various parametric perturbations (temperature, stress, and pressure). Surface acoustic wave (SAW) resonators are particularly well suited to such applications as their resonance frequency is directly influenced by these perturbations, modifying both the phase velocity and resonance conditions. Moreover, the intrinsic radio frequency (rf) nature of these devices makes them ideal for wireless applications, mainly exploiting antennas reciprocity and piezoelectric reversibility. In this paper, we present a wireless SAW sensor interrogation unit operating in the 434 MHz centered ISM band--selected as a tradeoff between antenna dimensions and electromagnetic wave penetration in dielectric media--based on the principles of a frequency sweep network analyzer. We particularly focus on the compliance with the ISM standard which reveals complicated by the need for switching from emission to reception modes similarly to radar operation. In this matter, we propose a fully digital rf synthesis chain to develop various interrogation strategies to overcome the corresponding difficulties and comply with the above-mentioned standard. We finally assess the reader interrogation range, accuracy, and dynamics.

Friedt, J.-M [SENSeOR, 32 Avenue de l'Observatoire, 25044 Besancon (France); Droit, C.; Martin, G.; Ballandras, S. [Department of Time and Frequency, FEMTO-ST, 32 Avenue de l'Observatoire, 25044 Besancon (France)

2010-01-15T23:59:59.000Z

182

A COMPLETE HISTORY OF THE HIGH-LEVEL WASTE PLANT AT THE WEST VALLEY DEMONSTRATION PROJECT  

SciTech Connect

The West Valley Demonstration Project (WVDP) vitrification melter was shut down in September 2002 after being used to vitrify High Level Waste (HLW) and process system residuals for six years. Processing of the HLW occurred from June 1996 through November 2001, followed by a program to flush the remaining HLW through to the melter. Glass removal and shutdown followed. The facility and process equipment is currently in a standby mode awaiting deactivation. During HLW processing operations, nearly 24 million curies of radioactive material were vitrified into 275 canisters of HLW glass. At least 99.7% of the curies in the HLW tanks at the WVDP were vitrified using the melter. Each canister of HLW holds approximately 2000 kilograms of glass with an average contact dose rate of over 2600 rem per hour. After vitrification processing ended, two more cans were filled using the Evacuated Canister Process to empty the melter at shutdown. This history briefly summarizes the initial stages of process development and earlier WVDP experience in the design and operation of the vitrification systems, followed by a more detailed discussion of equipment availability and failure rates during six years of operation. Lessons learned operating a system that continued to function beyond design expectations also are highlighted.

Petkus, Lawrence L.; Paul, James; Valenti, Paul J.; Houston, Helene; May, Joseph

2003-02-27T23:59:59.000Z

183

QUANTITY OF RADIATION REACHING GONADAL AREAS DURING THERAPY. IV. FACTORS INFLUENCING OVARY DOSE  

SciTech Connect

Attempts were made to evaluate the circumstances influencing the quantity of radiation reaching the ovaries during dermatologic radiation therapy, and to devise effective methods for reducing the amount to well within acceptable limits. Measurements were made in a specially constructed, life-size, pressed- wood (masonite) phantom which could be used in almost any position that might be assumed by man during routine x-ray therapy, and with provisions for insertion of an ionization chamber at the anatomic site representing an ovary. The various parameters which might influence ovary dose during conventional dermatologic x- ray procedures that were studied included: x-ray quality (kvp), tube current (ma), beam collimation and field size, shielding, angle of the beam in relation to the ovaries, and proximity of treatment site to the ovaries. Ovarian dose was measured during irradiation of the face, upper chest, and back, using each of the parameters alone and then in various combinations. The results, presented in tables and graphs, show that in order to minimize ovary dose during dermatologic x-ray therapy, one should utilize lower tube kilovoltage, softer radiation, appropriate collimation, effective shielding (minimizing area of field irraiated) with no added filtration, increased distance between the x-ray beam axis and the ovaries, and angling on the x-ray tube away from the ovaries. The gonad dose can best be reduced by exerting all of these simple and inexpensive means every time dermatologic radiation is administered. However, from these measurements it was evident that of the body areas studied, some may be irradiated without fear of exceeding even the max permissible dose to the ovaries, whereas other areas cannot be treated with x-rays without overdosing the ovaries regardless of the pre cautions taken. (BBB)

Witten, V.H.; Lee, H.

1963-05-01T23:59:59.000Z

184

Process and apparatus for determination and utilization of a quantity of preheating energy  

SciTech Connect

A preheating process and apparatus for determination and utilization of a quantity of preheating energy so as to automatically achieve a desired temperature at a given location prior to a predetermined period of usage, especially for preheating an automotive vehicle. In accordance with the process, the respective preheating times required to achieve a desired temperature for various possible actual conditions at the location of usage are determined and stored, at a predetermined point of time prior to the usage commencement time, the correct preheating time is determined from the stored preheating times based upon actual temperature conditions and application of heat to the location is automatically commenced at a time prior to the usage commencement time that corresponds to the difference between the usage commencement time and the correct preheating time determined previously. In accordance with a preferred embodiment of the apparatus, a preset-table timer actuated switching control means is provided for actuating a heating device. An actual temperature value sensor means is utilized to sense the actual temperature at the location to be preheated and is used to address a memory unit containing, as stored data, preheating times to be read out that correspond to the respectively required heating time to achieve a desired temperature for each of various possible actual temperatures at the location. The stored preheating time read out by the memory unit based upon actual temperature value sensed by the sensor means at a point of time prior to the preset usage commencement time is utilized by the switching control means to actuate the heating device at a time prior to a preset time for commencement of usage that corresponds to the difference between the preset usage commencement time and the time read out by the memory unit.

Lamkewitz, F.; Riedmaier, J.

1984-03-13T23:59:59.000Z

185

"Table A33. Total Quantity of Purchased Energy Sources by Census Region, Census Division,"  

U.S. Energy Information Administration (EIA) Indexed Site

Quantity of Purchased Energy Sources by Census Region, Census Division," Quantity of Purchased Energy Sources by Census Region, Census Division," " and Economic Characteristics of the Establishment, 1994" " (Estimates in Btu or Physical Units)" ,,,,,"Natural",,,"Coke" " ","Total","Electricity","Residual","Distillate","Gas(c)"," ","Coal","and Breeze","Other(d)","RSE" " ","(trillion","(million","Fuel Oil","Fuel Oil(b)","(billion","LPG","(1000 ","(1000","(trillion","Row" "Economic Characteristics(a)","Btu)","kWh)","(1000 bbl)","(1000 bbl)","cu ft)","(1000 bbl)","short tons)","short tons)","Btu)","Factors"

186

How parameters and regularization affect the Polyakov-Nambu-Jona-Lasinio model phase diagram and thermodynamic quantities  

SciTech Connect

We explore the phase diagram and the critical behavior of QCD thermodynamic quantities in the context of the so-called Polyakov-Nambu-Jona-Lasinio model. We show that this improved field theoretical model is a successful candidate for studying the equation of state and the critical behavior around the critical endpoint. We argue that a convenient choice of the model parameters is crucial to get the correct description of isentropic trajectories. The effects of the regularization procedure in several thermodynamic quantities is also analyzed. The results are compared with simple thermodynamic expectations and lattice data.

Costa, P. [Centro de Fisica Computacional, Departamento de Fisica, Universidade de Coimbra, P-3004-516 Coimbra (Portugal) and E.S.T.G., Instituto Politecnico de Leiria, Morro do Lena-Alto do Vieiro, 2411-901 Leiria (Portugal); Hansen, H. [IPNL, Universite de Lyon/Universite Lyon 1, CNRS/IN2P3, 4 rue E.Fermi, F-69622 Villeurbanne Cedex (France); Ruivo, M. C.; Sousa, C. A. de [Centro de Fisica Computacional, Departamento de Fisica, Universidade de Coimbra, P-3004-516 Coimbra (Portugal)

2010-01-01T23:59:59.000Z

187

A GREEN'S FUNCTION APPROACH FOR DETERMINING DOSE RATES FOR SMALL GRAM QUANTITIES IN SHIPPING PACKAGINGS  

Science Conference Proceedings (OSTI)

The Small Gram Quantity (SGQ) concept is based on the understanding that small amounts of hazardous materials, in this case radioactive materials (RAM), are significantly less hazardous than large amounts of the same materials. This paper describes a methodology designed to estimate an SGQ for several neutron and gamma emitting isotopes that can be shipped in a package in compliance with 10 CFR Part 71 external radiation level limits regulations. The neutron and photon sources were calculated using both ORIGEN-S and RASTA. The response from a unit source in each neutron and photon group was calculated using MCNP5 with each unshielded and shielded container configuration. Effects of self-shielding on both neutron and photon response were evaluated by including either plutonium oxide or iron in the source region for the case with no shielded container. For the cases of actinides mixed with light elements, beryllium is the bounding light element. The added beryllium (10 to 90 percent of the actinide mass) in the cases studied represents between 9 and 47 percent concentration of the total mixture mass. For beryllium concentrations larger than 50 percent, the increase in the neutron source term and dose rate tend to increase at a much lower rate than at concentrations lower than 50%. The intimately mixed actinide-beryllium form used in these models is very conservative and thus the limits presented in this report are practical bounds on the mass that can be safely shipped. The calculated dose rate from one gram of each isotope was then used to determin the maximum amount of a single isotope that could be shipped in the Model 9977 Package (or packagings having the same or larger external dimensions as well as similar structural materials) and have the external radiation level within the regulatory dose limits at the surface of the package. The estimates of the mass limits presented would also serve as conservative limits for both the Models 9975 and 9978 packages. If a package contains a mixture of isotopes, the acceptability for shipment can be determined by a sum of fractions approach. It should be noted that the SGQ masses presented in this report represent limits that would comply with the external radiation limits under 10CFR Part 71. They do not necessarily bound lower limits that may be required to comply with other factors such as heat load of the package.

Nathan, S.

2012-06-14T23:59:59.000Z

188

Preliminary Results of Voloxidation Processing of Kilogram Quantities of Used Nuclear Fuel  

SciTech Connect

Advanced nuclear fuel processing methodologies are being studied as part of the Advanced Fuel Cycle Initiative (AFCI) program at ORNL. To support this initiative, processes and equipment were deployed at ORNL to perform all steps in the recycle process on actual used nuclear fuels, ranging from used fuel receipt to production of products and waste forms at the kilogram-scale (with capacity to process 20 kg of used fuel per year in up to four campaigns). In the first campaign, approximately 4 kg of used fuel was processed. As previously reported, the head-end processing was completed using saw-segmented Dresden fuel in lab-scale equipment in multiple batches. The second processing campaign used a new single pin shear and a new bench-scale voloxidizer to perform the dry head-end treatment prior to fuel dissolution. Approximately ~5 kg of used fuel (heavy metal basis) was processed in the second campaign. Two different fuels were oxidized in three separate batches to provide a range of processing conditions. The material used for each batch and general processing conditions are summarized in Table 1. Progress of the oxidation reaction was monitored continuously by two primary measurements; the concentration of oxygen in the effluent stream which was depressed as the oxygen was consumed, and the concentration of krypton-85 in the effluent stream as measured by a gamma counter on the off-gas pipeline. Table 1. Voloxidation test conditions for second campaign. Batch Fuel Source Burnup (GWd/MT)Batch size (kg*)/(kg**)Segment Length (in) Oxidation GasOperation Temperature ( C) 1Surry-2361.223/1.7041.0Air500 2North Anna63 702.071/2.8850.88Air600 3North Anna63 702.012/2.8030.88Oxygen600 * Heavy metal basis. ** Total fuel (oxide + cladding) basis. Fission product gases evolved from the fuel during the oxidation process were trapped for subsequent chemical and radiochemical analysis. The series of traps included a bed of molecular sieves to recover tritium (as HTO), silver-substituted zeolite to capture iodine (e.g. as AgI), a caustic scrubber to collect carbon dioxide (including 14CO2), a hydrogen-substituted mordenite to capture krypton (e.g. 85Kr) by cryogenic temperature swing adsorption, and a silver-substituted mordenite to capture xenon by cryogenic temperature swing absorption. The quantities of these volatile gases collected were compared to ORIGEN calculations to estimate the effectiveness of the voloxidation process to separate the volatiles from the used fuel. This paper will describe the voloxidation system and present preliminary results from the second processing campaign.

Spencer, Barry B [ORNL; DelCul, Guillermo D [ORNL; Jubin, Robert Thomas [ORNL; Owens, R Steven [ORNL; Ramey, Dan W [ORNL; Collins, Emory D [ORNL

2009-01-01T23:59:59.000Z

189

A General, Cryogenically-Based Analytical Technique for the Determination of Trace Quantities of Volatile Organic Compounds in the Atmosphere  

Science Conference Proceedings (OSTI)

An analytical technique for the determination of trace (sub-ppbv) quantities of volatile organic compounds in air was developed. A liquid nitrogen-cooled trap operated at reduced pressures in series with a Dupont Nafion-based drying tube and a ...

Randolph A. Coleman; Wesley R. Cofer III; Robert A. Edahl Jr.

1985-09-01T23:59:59.000Z

190

Okanogan Subbasin Water Quality and Quantity Report for Anadromous Fish in 2006.  

DOE Green Energy (OSTI)

Fish need water of sufficient quality and quantity in order to survive and reproduce. The list of primary water quality indicators appropriate for monitoring of anadromous fish, as identified by the Upper Columbia Monitoring Strategy, includes: discharge, temperature, dissolved oxygen, pH, turbidity, conductivity, nitrogen, phosphorus and ammonia. The Colville Tribes Fish and Wildlife Department began evaluating these water quality indicators in 2005 and this report represents data collected from October 1, 2005 through September 30, 2006. We collected empirical status and trend data from various sources to evaluate each water quality indicator along the main stem Okanogan and Similkameen Rivers along with several tributary streams. Each water quality indicator was evaluated based upon potential impacts to salmonid survival or productivity. Specific conductance levels and all nutrient indicators remained at levels acceptable for growth, survival, and reproduction of salmon and steelhead. These indicators were also considered of marginal value for monitoring environmental conditions related to salmonids within the Okanogan subbasin. However, discharge, temperature, turbidity, dissolved oxygen and pH in that order represent the water quality indicators that are most useful for monitoring watershed health and habitat changes and will help to evaluate threats or changes related to salmon and steelhead restoration and recovery. On the Okanogan River minimum flows have decreased over the last 12 years at a rate of -28.3CFS/year as measured near the town of Malott, WA. This trend is not beneficial for salmonid production and efforts to reverse this trend should be strongly encouraged. Turbidity levels in Bonaparte and Omak Creek were a concern because they had the highest monthly average readings. Major upland disturbance in the Bonaparte Creek watershed has occurred for decades and agricultural practices within the riparian areas along this creek have lead to major channel incision and bank instability. High sediment loads continue to threaten Omak and Bonaparte sub-watersheds. Major rehabilitation efforts are needed within these sub-watersheds to improve salmonid habitats. We found that for the past 12 years dissolved oxygen levels have been on a slightly downward trend during summer/fall Chinook egg incubation. Dissolved oxygen readings in early October, for summer/fall Chinook and from June through early July for summer steelhead can occasionally drop to the range from 8 to 10 mg/L and therefore warrant continued monitoring. Levels of pH represent an indicator that has little monitoring value throughout most of the subbasin. The Similkameen River drainage showed dramatic annual changes in the mean pH values and a declining trend for pH thus warranting continued monitoring. Average daily temperatures, in 2006, exceeded 25 C for eight days in July in the Okanogan River at Malott. Due to increased warm water temperatures, delays in migration have increased at a rate of 1.82 days per year over the last 10 years. Increases in water temperature can be linked to many anthropogenic activities. Increasing water temperatures within the Okanogan River watershed represent the single most limiting factor facing salmonids in main-stem habitats.

Colville Tribes, Department of Fish & Wildlife

2007-12-01T23:59:59.000Z

191

Boehmite Actual Waste Dissolutions Studies  

SciTech Connect

The U.S. Department of Energy plans to vitrify approximately 60,000 metric tons of high-level waste (HLW) sludge from underground storage tanks at the Hanford Nuclear Reservation. To reduce the volume of HLW requiring treatment, a goal has been set to remove a significant quantity of the aluminum, which comprises nearly 70 percent of the sludge. Aluminum is found in the form of gibbsite, sodium aluminate and boehmite. Gibbsite and sodium aluminate can be easily dissolved by washing the waste stream with caustic. Boehmite, which comprises nearly half of the total aluminum, is more resistant to caustic dissolution and requires higher treatment temperatures and hydroxide concentrations. Samples were taken from four Hanford tanks and homogenized in order to give a sample that is representative of REDOX (Reduction Oxidation process for Pu recovery) sludge solids. Bench scale testing was performed on the homogenized waste to study the dissolution of boehmite. Dissolution was studied at three different hydroxide concentrations, with each concentration being run at three different temperatures. Samples were taken periodically over the 170 hour runs in order to determine leaching kinetics. Results of the dissolution studies and implications for the proposed processing of these wastes will be discussed.

Snow, Lanee A.; Lumetta, Gregg J.; Fiskum, Sandra K.; Peterson, Reid A.

2008-07-15T23:59:59.000Z

192

A Little Here, A Little There, A Fairly Big Problem Everywhere: Small Quantity Site Transuranic Waste Disposition Alternatives  

Science Conference Proceedings (OSTI)

Small quantities of transuranic (TRU) waste represent a significant challenge to the waste disposition and facility closure plans of several sites in the Department of Energy (DOE) complex. This paper presents the results of a series of evaluations, using a systems engineering approach, to identify the preferred alternative for dispositioning TRU waste from small quantity sites (SQSs). The TRU waste disposition alternatives evaluation used semi-quantitative data provided by the SQSs, potential receiving sites, and the Waste Isolation Pilot Plant (WIPP) to select and recommend candidate sites for waste receipt, interim storage, processing, and preparation for final disposition of contact-handled (CH) and remote-handled (RH) TRU waste. The evaluations of only four of these SQSs resulted in potential savings to the taxpayer of $33 million to $81 million, depending on whether mobile systems could be used to characterize, package, and certify the waste or whether each site would be required to perform this work. Small quantity shipping sites included in the evaluation included the Battelle Columbus Laboratory (BCL), University of Missouri Research Reactor (MURR), Energy Technology Engineering Center (ETEC), and Mound. Candidate receiving sites included the Idaho National Engineering and Environmental Laboratory (INEEL), the Savannah River Site (SRS), Los Alamos National Laboratory (LANL), Oak Ridge (OR), and Hanford. At least 14 additional DOE sites having TRU waste may be able to save significant money if cost savings are similar to the four evaluated thus far.

Luke, Dale Elden; Parker, Douglas Wayne; Moss, J.; Monk, Thomas Hugh; Fritz, Lori Lee; Daugherty, B.; Hladek, K.; Kosiewicx, S.

2000-03-01T23:59:59.000Z

193

A little here, a little there, a fairly big problem everywhere: Small quantity site transuranic waste disposition alternatives  

Science Conference Proceedings (OSTI)

Small quantities of transuranic (TRU) waste represent a significant challenge to the waste disposition and facility closure plans of several sites in the Department of Energy (DOE) complex. This paper presents the results of a series of evaluations, using a systems engineering approach, to identify the preferred alternative for dispositioning TRU waste from small quantity sites (SQSs). The TRU waste disposition alternatives evaluation used semi-quantitative data provided by the SQSs, potential receiving sites, and the Waste Isolation Pilot Plant (WIPP) to select and recommend candidate sites for waste receipt, interim storage, processing, and preparation for final disposition of contact-handled (CH) and remote-handled (RH) TRU waste. The evaluations of only four of these SQSs resulted in potential savings to the taxpayer of $33 million to $81 million, depending on whether mobile systems could be used to characterize, package, and certify the waste or whether each site would be required to perform this work. Small quantity shipping sites included in the evaluation included the Battelle Columbus Laboratory (BCL), University of Missouri Research Reactor (MURR), Energy Technology Engineering Center (ETEC), and Mound Laboratory. Candidate receiving sites included the Idaho National Engineering and Environmental Laboratory (INEEL), the Savannah River Site (SRS), Los Alamos National Laboratory (LANL), Oak Ridge (OR), and Hanford. At least 14 additional DOE sites having TRU waste may be able to save significant money if cost savings are similar to the four evaluated thus far.

D. Luke; D. Parker; J. Moss; T. Monk (INEEL); L. Fritz (DOE-ID); B. Daugherty (SRS); K. Hladek (WM Federal Services Hanford); S. Kosiewicx (LANL)

2000-02-27T23:59:59.000Z

194

Determining the quality and quantity of heat produced by proton exchange membrane fuel cells with application to air-cooled stacks for combined heat and power  

E-Print Network (OSTI)

Determining the quality and quantity of heat produced by proton exchange membrane fuel cells Determining the quality and quantity of heat produced by proton exchange membrane fuel cells with application, the coolant is pumped to a heat recovery system. A water-to-air heat exchange system or water-to-water heat

Victoria, University of

195

Economic Analysis of Energy Crop Production in the U.S. - Location, Quantities, Price, and Impacts on Traditional Agricultural Crops  

DOE Green Energy (OSTI)

POLYSYS is used to estimate US locations where, for any given energy crop price, energy crop production can be economically competitive with conventional crops. POLYSYS is a multi-crop, multi-sector agricultural model developed and maintained by the University of Tennessee and used by the USDA-Economic Research Service. It includes 305 agricultural statistical districts (ASD) which can be aggregated to provide state, regional, and national information. POLYSYS is being modified to include switchgrass, hybrid poplar, and willow on all land suitable for their production. This paper summarizes the preliminary national level results of the POLYSYS analysis for selected energy crop prices for the year 2007 and presents the corresponding maps (for the same prices) of energy crop production locations by ASD. Summarized results include: (1) estimates of energy crop hectares (acres) and quantities (dry Mg, dry tons), (2) identification of traditional crops allocated to energy crop production and calculation of changes in their prices and hectares (acres) of production, and (3) changes in total net farm returns for traditional agricultural crops. The information is useful for identifying areas of the US where large quantities of lowest cost energy crops can most likely be produced.

Walsh, M.E.; De La Torre Ugarte, D.; Slinsky, S.; Graham, R.L.; Shapouri, H.; Ray, D.

1998-10-04T23:59:59.000Z

196

Definition of Small Gram Quantity Contents for Type B Radioactive Material Transportation Packages: Activity-Based Content Limitations  

SciTech Connect

Since the 1960's, the Department of Transportation Specification (DOT Spec) 6M packages have been used extensively for transportation of Type B quantities of radioactive materials between Department of Energy (DOE) facilities, laboratories, and productions sites. However, due to the advancement of packaging technology, the aging of the 6M packages, and variability in the quality of the packages, the DOT implemented a phased elimination of the 6M specification packages (and other DOT Spec packages) in favor of packages certified to meet federal performance requirements. DOT issued the final rule in the Federal Register on October 1, 2004 requiring that use of the DOT Specification 6M be discontinued as of October 1, 2008. A main driver for the change was the fact that the 6M specification packagings were not supported by a Safety Analysis Report for Packaging (SARP) that was compliant with Title 10 of the Code of Federal Regulations part 71 (10 CFR 71). Therefore, materials that would have historically been shipped in 6M packages are being identified as contents in Type B (and sometimes Type A fissile) package applications and addenda that are to be certified under the requirements of 10 CFR 71. The requirements in 10 CFR 71 include that the Safety Analysis Report for Packaging (SARP) must identify the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents (10 CFR 71.33(b)(1) and 10 CFR 71.33(b)(2)), and that the application (i.e., SARP submittal or SARP addendum) demonstrates that the external dose rate (due to the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents) on the surface of the packaging (i.e., package and contents) not exceed 200 mrem/hr (10 CFR 71.35(a), 10 CFR 71.47(a)). It has been proposed that a 'Small Gram Quantity' of radioactive material be defined, such that, when loaded in a transportation package, the dose rates at external points of an unshielded packaging not exceed the regulatory limits prescribed by 10 CFR 71 for non-exclusive shipments. The mass of each radioisotope presented in this paper is limited by the radiation dose rate on the external surface of the package, which per the regulatory limit should not exceed 200 mrem/hr. The results presented are a compendium of allowable masses of a variety of different isotopes (with varying impurity levels of beryllium in some of the actinide isotopes) that, when loaded in an unshielded packaging, do not result in an external dose rate on the surface of the package that exceeds 190 mrem/hr (190 mrem/hr was chosen to provide 5% conservatism relative to the regulatory limit). These mass limits define the term 'Small Gram Quantity' (SGQ) contents in the context of radioactive material transportation packages. The term SGQ is isotope-specific and pertains to contents in radioactive material transportation packages that do not require shielding and still satisfy the external dose rate requirements. Since these calculated mass limits are for contents without shielding, they are conservative for packaging materials that provide some limited shielding or if the contents are placed into a shielded package. The isotopes presented in this paper were chosen as the isotopes that Department of Energy (DOE) sites most likely need to ship. Other more rarely shipped isotopes, along with industrial and medical isotopes, are planned to be included in subsequent extensions of this work.

Sitaraman, S; Kim, S; Biswas, D; Hafner, R; Anderson, B

2010-10-27T23:59:59.000Z

197

Non-Gaussianity effect of petrophysical quantities by using q-entropy and multi fractal random walk  

E-Print Network (OSTI)

The geological systems such as petroleum reservoirs is investigated by the entropy introduced by Tsallis and multiplicative hierarchical cascade model. When non-Gaussianity appears, it is sign of uncertainty and phase transition, which could be sign of existence of petroleum reservoirs. Two important parameters which describe a system at any scale are determined; the non-Gaussian degree, $q$, announced in entropy and the intermittency, $\\lambda^2$, which explains a critical behavior in the system. There exist some petrophysical indicators in order to characterize a reservoir, but there is vacancy to measure scaling information contain in comparison with together, yet. In this article, we compare the non-Gaussianity in three selected indicators in various scales. The quantities investigated in this article includes Gamma emission (GR), sonic transient time (DT) and Neutron porosity (NPHI). It is observed that GR has a fat tailed PDF at all scales which is a sign of phase transition in the system which indicate...

lai, Z Koohi; Jafari, G R

2013-01-01T23:59:59.000Z

198

The Ratio of Diffuse to Direct Solar Irradiance (Perpendicular to the Sun's Rays) with Clear SkiesA Conserved Quantity Throughout the Day  

Science Conference Proceedings (OSTI)

The ratio of diffuse irradiance to direct normal is a conserved quantity throughout the day. Though its absolute value depends on the condition of the atmosphere, ground reflection and obstruction of horizon, once this value is established (by a ...

William A. Peterson; Inge Dirmhirn

1981-07-01T23:59:59.000Z

199

Ethanol and Its Effect on the U.S. Corn Market: How the Price of E-85 Influences Equilibrium Corn Prices and Equilibrium Quantity.  

E-Print Network (OSTI)

??This study analyzes the impact the market price of E-85 has on equilibrium price and quantity exchanged of corn in the U.S. market. After presenting (more)

PINCIN, JARED

2007-01-01T23:59:59.000Z

200

Techniques of evaluation of QCD low-energy physical quantities with running coupling with infrared fixed point  

E-Print Network (OSTI)

Perturbative QCD (pQCD) running coupling a(Q^2) (=alpha_s(Q^2)/pi) is expected to get modified at low spacelike momenta 0 1 GeV by nonperturbative (NP) terms, typically by some power-suppressed terms ~1/(Q^2)^N. Evaluations of low-energy physical QCD quantities in terms of such A(Q^2) couplings (with IR fixed point) at a level beyond one-loop are usually performed with (truncated) power series in A(Q^2). We argue that such an evaluation is not correct, because the NP terms in general get out of control as the number of terms in the power series increases. The series consequently become increasingly unstable under the variation of the renormalization scale, and have a fast asymptotic divergent behavior compounded by the renormalon problem. We argue that an alternative series in terms of logarithmic derivatives of A(Q^2) should be used. Further, a Pad\\'e-related resummation based on this series gives results which are renormalization scale independent and show very good convergence. Timelike low-energy observables can be evaluated analogously, using the integral transformation which relates the timelike observable with the corresponding spacelike observable.

Gorazd Cveti?

2013-09-06T23:59:59.000Z

Note: This page contains sample records for the topic "vitrified hlw quantities" from the National Library of EnergyBeta (NLEBeta).
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201

An indirect sensing technique for diesel fuel quantity control. Technical progress report, October 1--December 31, 1998  

DOE Green Energy (OSTI)

Work has proceeded intensely with the objective of completing the commercial prototype system prior to the end of the contract period. At the time of this report, testing and refinement of the commercial version of the system has not been completed. During this reporting period, several major milestones were reached and many significant lessons were learned. These are described. The experimental retrofit system has achieved all performance objectives in engine dynamometer tests. The prototype commercial version of the system will begin demonstration service on the first of several Santa Maria Area Transit (SMAT) transit buses on February 1, 1999. The commercial system has been redesignated the Electronic Diesel Smoke Reduction System (EDSRS) replacing the original internal pseudonym ADSC. The focus has been narrowed to a retrofit product suitable for installation on existing mechanically-governed diesel engines. Included in this potential market are almost all diesel-powered passenger cars and light trucks manufactured prior to the introduction of the most recent clean diesel engines equipped with particulate traps and electronic controls. Also included are heavy-duty trucks, transit vehicles, school buses, and agricultural equipment. This system is intended to prevent existing diesel engines from overfueling to the point of visible particulate emissions (smoke), while allowing maximum smoke-limited torque under all operating conditions. The system employs a microcontroller and a specialized exhaust particulate emission sensor to regulate the maximum allowable fuel quantity via an adaptive throttle-limit map. This map specifies a maximum allowable throttle position as a function of engine speed, turbocharger boost pressure and engine coolant temperature. The throttle position limit is mechanized via a servo actuator inserted in the throttle cable leading to the injection pump.

MacCarley, C.A.

1999-01-26T23:59:59.000Z

202

A solid state approach to the production of kilogram quantities of Si[sub 80]Ge[sub 20] thermoelectric alloys  

DOE Green Energy (OSTI)

An important consideration in the development of improved materials for thermal-to-electrical power generation is whether a research-scale process or methodology is amenable to production of kilogram quantities. Research efforts on the solid state technique of mechanical alloying have shown that both n- and p-type Si-20 at. % Ge alloys can be produced which have improved thermoelectric properties compared to state-of-the-art MOD-RTG materials. Studies on the production of large quantities of mechanically alloyed powder alloys using a planetary mill indicate that properties similar to those observed in alloys prepared in smaller quantities by a vibratory mill can be obtained. The characterization of several p-type alloys doped with 0.8 at. % B in the form SiB[sub 4] by X-ray diffraction, scanning laser mass spectroscopy, Hall effect, and high temperature electrical resistivity and Seebeck coefficient measurements are described. The transport properties of these alloys are shown to be comparable to those measured on similar samples prepared in small quantities by a research-grade vibratory mill.

Cook, B.A.; Harringa, J.L.; Beaudry, B.J.

1992-01-01T23:59:59.000Z

203

PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION  

Science Conference Proceedings (OSTI)

In accordance with the Nuclear Waste Policy Amendments Act of 1987, Yucca Mountain was designated as the site to be investigated as a potential repository for the disposal of high-level radioactive waste. The Yucca Mountain site is an undeveloped area located on the southwestern edge of the Nevada Test Site (NTS), about 100 miles northwest of Las Vegas. The site currently lacks rail service or an existing right-of-way. If the Yucca Mountain site is found suitable for the repository, rail service is desirable to the Office of Civilian Waste Management (OCRWM) Program because of the potential of rail transportation to reduce costs and to reduce the number of shipments relative to highway transportation. A Preliminary Rail Access Study evaluated 13 potential rail spur options. Alternative routes within the major options were also developed. Each of these options was then evaluated for potential land use conflicts and access to regional rail carriers. Three potential routes having few land use conflicts and having access to regional carriers were recommended for further investigation. Figure 1-1 shows these three routes. The Jean route is estimated to be about 120 miles long, the Carlin route to be about 365 miles long, and Caliente route to be about 365 miles long. The remaining ten routes continue to be monitored and should any of the present conflicts change, a re-evaluation of that route will be made. Complete details of the evaluation of the 13 routes can be found in the previous study. The DOE has not identified any preferred route and recognizes that the transportation issues need a full and open treatment under the National Environmental Policy Act. The issue of transportation will be included in public hearings to support development of the Environmental Impact Statement (EIS) proceedings for either the Monitored Retrievable Storage Facility or the Yucca Mountain Project or both.

D.C. Richardson

2003-03-19T23:59:59.000Z

204

Thermodynamic quantities and defect equilibrium in La{sub 2-x}Sr{sub x}NiO{sub 4+{delta}}  

Science Conference Proceedings (OSTI)

In order to elucidate the relation between thermodynamic quantities, the defect structure, and the defect equilibrium in La{sub 2-x}Sr{sub x}NiO{sub 4+{delta}}, statistical thermodynamic calculation is carried out and calculated results are compared to those obtained from experimental data. Partial molar enthalpy of oxygen and partial molar entropy of oxygen are obtained from delta-P(O{sub 2})-T relation by using Gibbs-Helmholtz equation. Statistical thermodynamic model is derived from defect equilibrium models proposed before by authors, localized electron model and delocalized electron model which could well explain the variation of oxygen content of La{sub 2-x}Sr{sub x}NiO{sub 4+{delta}}. Although assumed defect species and their equilibrium are different, the results of thermodynamic calculation by localized electron model and delocalized electron model show minor difference. Calculated results by the both models agree with the thermodynamic quantities obtained from oxygen nonstoichiometry of La{sub 2-x}Sr{sub x}NiO{sub 4+{delta}}. - Graphical abstract: In order to elucidate the relation between thermodynamic quantities, the defect structure, and the defect equilibrium in La{sub 2-x}Sr{sub x}NiO{sub 4+{delta}}, statistics thermodynamic calculation is carried out and calculated results are compared to those obtained from experimental data.

Nakamura, Takashi, E-mail: t-naka@mail.tagen.tohoku.ac.j [Institute of Multidisciplinary Research for Advanced Materials, Tohoku University, 2-1-1, Katahira, Aoba-Ku, Sendai 980-8577 (Japan); Yashiro, Keiji; Sato, Kazuhisa; Mizusaki, Junichiro [Institute of Multidisciplinary Research for Advanced Materials, Tohoku University, 2-1-1, Katahira, Aoba-Ku, Sendai 980-8577 (Japan)

2009-05-15T23:59:59.000Z

205

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

SciTech Connect

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

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

1999-05-01T23:59:59.000Z

206

Hanford Tank Waste What is in it? Where is it going?  

Science Conference Proceedings (OSTI)

The Waste Treatment Plant currently under construction for treatment of High Level Waste at the Hanford Site will process High Level Waste (HLW) to reduce the quantity of HLW material that must be immobilized. Recently, an extensive testing program was undertaken to characterize the composition of some of the major sources of HLW in the Hanford tank farm system. This effort has led to an increased understanding of the chemical form and the underlying dissolution chemistry for much of the waste.

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

2010-11-30T23:59:59.000Z

207

H:\cindy_pratt\hlw rod.tif  

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

RECORD OF DECISION RECORD OF DECISION For The Idaho High- Level Waste and Facilities Disposition Final Environmental Impact Statement December 2005 United States Department of Energy 1 U.S. DEPARTMENT OF ENERGY Office of Environmental Management Record of Decision for the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement SUMMARY: DOE is making decisions pursuant to the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (Final EIS) (DOEÆIS-287), issued in October 2002. The Final EIS presents the analysis of a proposed action containing two sets of alternatives: (1) waste processing alternatives for treating, storing and disposing of liquid mixed (radioactive and hazardous) transuranic (TRU) waste/sodium-bearing

208

Waste Treatment and Immobilation Plant HLW Waste Vitrification...  

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

compounds VSL Vitreous State Laboratory of the Catholic University of America WESP Wet Electrostatic Precipitator WGI Washington Group International WTP Waste Treatment and...

209

Molybdenum Oxide Behavior in French HLW Nuclear Glass: Current ...  

Science Conference Proceedings (OSTI)

Rare Earth and Plutonium Doping of Apatite Secure and Certify Studies to Work on Production of Spiked Plutonium Vitrification of DOE Problematic Wastes...

210

Property Models for High Waste Loaded Hanford HLW Glasses  

High Waste Loading Was Shown for Selected Wastes Examples of the high loaded glasses Al 2O 3 loadings in the 24-26 wt% range compared to <15% for a

211

Is Buoyancy a Relative Quantity?  

Science Conference Proceedings (OSTI)

Basic concepts of buoyancy are reviewed and considered first in light of simple parcel theory and then in a more complete form. It is shown that parcel theory is generally developed in terms of the density (temperature) difference between an ...

Charles A. Doswell III; Paul M. Markowski

2004-04-01T23:59:59.000Z

212

Uncertainties in parton related quantities.  

E-Print Network (OSTI)

of the form ?ns lnn?1(1/x) and ?ns ln2n?1(1? x) in the perturbative expansion. This means that renormalization and factorization scale variation are not a reliable way of estimating higher order effects, e.g., at small x P 1qg ? ?S(2) P 2qg ? ?s(2) x (28... ) whereas Pnqg ? ?nS(2) lnn?2(1/x) x . (29) and scale variations of P 1qg, P 2qg never give an indication of these terms. Hence, in order to investigate the true theoretical error we must consider some way of performing correct large and small x...

Thorne, Robert S

213

Method for quantitative determination and separation of trace amounts of chemical elements in the presence of large quantities of other elements having the same atomic mass  

DOE Patents (OSTI)

Photoionization via autoionizing atomic levels combined with conventional mass spectroscopy provides a technique for quantitative analysis of trace quantities of chemical elements in the presence of much larger amounts of other elements with substantially the same atomic mass. Ytterbium samples smaller than 10 ng have been detected using an ArF* excimer laser which provides the atomic ions for a time-of-flight mass spectrometer. Elemental selectivity of greater than 5:1 with respect to lutetium impurity has been obtained. Autoionization via a single photon process permits greater photon utilization efficiency because of its greater absorption cross section than bound-free transitions, while maintaining sufficient spectroscopic structure to allow significant photoionization selectivity between different atomic species. Separation of atomic species from others of substantially the same atomic mass is also described.

Miller, C.M.; Nogar, N.S.

1982-09-02T23:59:59.000Z

214

Numerical evidence for relevance of disorder in a Poland-Scheraga DNA denaturation model with self-avoidance: Scaling behavior of average quantities  

E-Print Network (OSTI)

We study numerically the effect of sequence heterogeneity on the thermodynamic properties of a Poland-Scheraga model for DNA denaturation taking into account self-avoidance, i.e. with exponent c_p=2.15 for the loop length probability distribution. In complement to previous on-lattice Monte Carlo like studies, we consider here off-lattice numerical calculations for large sequence lengths, relying on efficient algorithmic methods. We investigate finite size effects with the definition of an appropriate intrinsic length scale x, depending on the parameters of the model. Based on the occurrence of large enough rare regions, for a given sequence length N, this study provides a qualitative picture for the finite size behavior, suggesting that the effect of disorder could be sensed only with sequence lengths diverging exponentially with x. We further look in detail at average quantities for the particular case x=1.3, ensuring through this parameter choice the correspondence between the off-lattice and the on-lattice studies. Taken together, the various results can be cast in a coherent picture with a crossover between a nearly pure system like behavior for small sizes N = 2/d (=2).

Barbara Coluzzi; Edouard Yeramian

2006-11-28T23:59:59.000Z

215

Seagate Crystal Reports - sum5.  

Office of Environmental Management (EM)

Activities for the Current Year (Sum-5) Total In Situ M a n a g e d Quantity (m 3) W a s t e T y p e T r e a t m e n t Q u a n i t y ( m 3 ) Dis p o s a l Q u a n t i t y ( m 3 ) O t h e r P r o c e s s i n g Quantity (m 3) Current Year: 2000 0.00 6,179.16 0.00 13,302.16 High Level Waste 7,123.00 0.00 0.00 0.00 0.00 HLW - Vitrified* 0.00 25,224,051.67 109,482.56 620,184.69 25,997,927.64 Low Level Waste 44,208.72 1,769,462.20 26,793.32 12,478.78 1,813,177.23 Mixed Low Level Waste 4,442.93 290,997.70 89.16 734.39 296,005.30 Transuranic Waste 4,184.05 145,882.00 0.00 183,013.00 333,902.56 Other** 5,007.56 O n S i t e T r e a t m e n t ( M T H M ) S h i p t o O t h e r D O E S i t e s f o r M anagam e n t /Storage Ship for Final Dis p o s a l ( M T H M ) Total (M T HM ) 14.6110 14.5110 0.1000 0.0000 Spent Nuclear Fuel*** Data Set ID: EM Corporate - FY 2001 Update

216

Permitting plan for the high-level waste interim storage  

SciTech Connect

This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist.

Deffenbaugh, M.L.

1997-04-23T23:59:59.000Z

217

Method for selective recovery of PET-usable quantities of [.sup.18 F] fluoride and [.sup.13 N] nitrate/nitrite from a single irradiation of low-enriched [.sup.18 O] water  

DOE Patents (OSTI)

A process for simultaneously producing PET-usable quantities of [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- for radiotracer synthesis is disclosed. The process includes producing [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [.sup.18 F]F.sup.- simultaneously by exposing a low-enriched (20%-30%) [.sup.18 O]H.sub.2 O target to proton irradiation, sequentially isolating the [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [.sup.18 F]F.sup.- from the [.sup.18 O]H.sub.2 O target, and reducing the [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- to [.sup.13 N]NH.sub.3. The [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- products are then conveyed to a laboratory for radiotracer applications. The process employs an anion exchange resin for isolation of the isotopes from the [.sup.18 O]H.sub.2 O, and sequential elution of [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [ .sup.18 F]F.sup.- fractions. Also the apparatus is disclosed for simultaneously producing PET-usable quantities of [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- from a single irradiation of a single low-enriched [.sup.18 O]H.sub.2 O target.

Ferrieri, Richard A. (Patchogue, NY); Schlyer, David J. (Bellport, NY); Shea, Colleen (Wading River, NY)

1995-06-13T23:59:59.000Z

218

Summary - Demonstration Bulk Vitrification System (DBVS) for Low-Actvity Waste at Hanford  

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

DBVS DBVS ETR Report Date: September 2006 ETR-3 United States Department of Energy Office of Environmental Management (DOE-EM) External Technical Review of the Demonstration Bulk Vitrification System (DBVS) for Low Activity Waste (LAW) at Hanford Why DOE-EM Did This Review The Department of Energy (DOE) is charged with the safe retrieval, treatment and disposal of 53 million gallons of Hanford radioactive waste. The Waste Treatment Plant (WTP) is being designed to treat and vitrify the High Level Waste (HLW) fraction in 20-25 years. The WTP is undersized for vitrifying the LAW fraction over the same time frame. The DOE is evaluating Bulk Vitrification as an alternative to increasing the size of the WTP LAW treatment process. Bulk vitrification is an in-container melting

219

EIS-0082-S2: Savannah River Site Salt Processing, Savannah River Site,  

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

082-S2: Savannah River Site Salt Processing, Savannah River 082-S2: Savannah River Site Salt Processing, Savannah River Site, Aiken, South Carolina EIS-0082-S2: Savannah River Site Salt Processing, Savannah River Site, Aiken, South Carolina SUMMARY This SEIS evaluates the potential environmental impacts of alternatives for separating the high-activity fraction from the low-activity fraction of the high-level radioactive waste salt solutions now stored in underground tanks at the Savannah River Site (SRS) near Aiken, South Carolina. The high-activity fraction of the high-level waste (HLW) salt solution would then be vitrified in the Defense Waste Processing Facility (DWPF) and stored until it could be disposed of as HLW in a geologic repository. The low activity fraction would be disposed of as low-level waste (saltstone)

220

EIS-0082-S2: Final Supplemental Environmental Impact Statement | Department  

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

Final Supplemental Environmental Impact Statement Final Supplemental Environmental Impact Statement EIS-0082-S2: Final Supplemental Environmental Impact Statement Savannah River Site Salt Processing This SEIS evaluates the potential environmental impacts of alternatives for separating the high-activity fraction from the low-activity fraction of the high-level radioactive waste salt solutions now stored in underground tanks at the Savannah River Site (SRS) near Aiken, South Carolina. The high-activity fraction of the high-level waste (HLW) salt solution would then be vitrified in the Defense Waste Processing Facility (DWPF) and stored until it could be disposed of as HLW in a geologic repository. The low activity fraction would be disposed of as low-level waste (saltstone) in vaults at SRS. EIS-0082-S2-2001.pdf

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221

High-Level Waste Melter Review  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy (DOE) is faced with a massive cleanup task in resolving the legacy of environmental problems from years of manufacturing nuclear weapons. One of the major activities within this task is the treatment and disposal of the extremely large amount of high-level radioactive (HLW) waste stored at the Hanford Site in Richland, Washington. The current planning for the method of choice for accomplishing this task is to vitrify (glassify) this waste for disposal in a geologic repository. This paper describes the results of the DOE-chartered independent review of alternatives for solidification of Hanford HLW that could achieve major cost reductions with reasonable long-term risks, including recommendations on a path forward for advanced melter and waste form material research and development. The potential for improved cost performance was considered to depend largely on increased waste loading (fewer high-level waste canisters for disposal), higher throughput, or decreased vitrification facility size.

Ahearne, J.; Gentilucci, J.; Pye, L. D.; Weber, T.; Woolley, F.; Machara, N. P.; Gerdes, K.; Cooley, C.

2002-02-26T23:59:59.000Z

222

Publication Price Quantity Total Economic Outlook Studies  

E-Print Network (OSTI)

to Work in WV 2010 20.00$ $ Other Studies >> The Economic Impact of the Natural Gas Industry and the Marcellus Shale Development in West Virginia in 2009 20.00$ ___________ >> Consensus Coal Production

Mohaghegh, Shahab

223

What are vector quantities and how do...  

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

science books, they usually stick to 2 dimensional problems (i.e., the surface of the book page) for simplicity. Vectors are shown as a line segment with an arrow head at one...

224

Thermodynamic Quantities for the Ionization Reactions of ...  

Science Conference Proceedings (OSTI)

... CAPS in that all three substances are sulfonic acids that differ ... Other names acetic acid; athylic acid; glacial acetic acid; vinegar; methanecarboxylic ...

2008-06-09T23:59:59.000Z

225

Underground storage tank integrated demonstration: Evaluation of pretreatment options for Hanford tank wastes  

SciTech Connect

Separation science plays a central role inn the pretreatment and disposal of nuclear wastes. The potential benefits of applying chemical separations in the pretreatment of the radioactive wastes stored at the various US Department of Energy sites cover both economic and environmental incentives. This is especially true at the Hanford Site, where the huge volume (>60 Mgal) of radioactive wastes stored in underground tanks could be partitioned into a very small volume of high-level waste (HLW) and a relatively large volume of low-level waste (LLW). The cost associated with vitrifying and disposing of just the HLW fraction in a geologic repository would be much less than those associated with vitrifying and disposing of all the wastes directly. Futhermore, the quality of the LLW form (e.g., grout) would be improved due to the lower inventory of radionuclides present in the LLW stream. In this report, we present the results of an evaluation of the pretreatment options for sludge taken from two different single-shell tanks at the Hanford Site-Tanks 241-B-110 and 241-U-110 (referred to as B-110 and U-110, respectively). The pretreatment options examined for these wastes included (1) leaching of transuranic (TRU) elements from the sludge, and (2) dissolution of the sludge followed by extraction of TRUs and {sup 90}Sr. In addition, the TRU leaching approach was examined for a third tank waste type, neutralized cladding removal waste.

Lumetta, G.J.; Wagner, M.J.; Colton, N.G.; Jones, E.O.

1993-06-01T23:59:59.000Z

226

Microsoft PowerPoint - 9-07 Case Atlanta HLW techexchange.ppt  

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

Calcine Disposition Project Calcine Disposition Project Joel Case Calcine Disposition Project Federal Project Director November 17, 2010 Print Close 2 2 Calcine is Solidified First & Other Cycle Raffinates * Resulted in a 7 to 1 volume reduction * Capable of being safely stored for several hundred years in 43 large shielded bins contained in six bin sets Print Close 3 3 Calcine Generation History Campaign Parameters Begin End Volume gal Volume ft 3 Volume m 3 Curies Dec- 63 Mar - 81 4,081,000 77,300 2,189 Aug - 82 May - 00 3,644,000 78,000 2,209 Total 7,725,000 155,300 4,398 3.64E+07 Print Close 4 4 Calcine Solids Storage Facility (CSSF) Status CSSF Bins Capacity (m 3 ) In Use 1 12 227 2 7 851 3 7 1,130 4 3 486 5 7 1,010 6 7 1,506 Sub 43 5,210 Not in Use 7 7 1,784 Total 50 6,994 1 2 3 4 5 6 Print Close 5 Hot Isostatic Press Treatment Technology * The Department Selected Hot Isostatic

227

Investigation of Cold Cap Behavior in HLW Melter through an Array ...  

Science Conference Proceedings (OSTI)

Symposium, Materials Issues in Nuclear Waste Management in the 21st Century ... the batch-to-glass conversion as it occurs in high-level-waste glass processing melters. ... The Properties of Spent Nuclear Fuel under Waste Disposal Conditions ... UK Radioactive Waste: Classification, Sources and Management Strategies.

228

Physical, Chemical and Structural Evolution of Zeolite - Containing Waste Forms Produced from Metakaolinite and Calcined HLW  

Science Conference Proceedings (OSTI)

During the seventh year of the current grant (DE-FG02-05ER63966) we completed an exhaustive study of cold calcination and began work on the development of tank fill materials to fill empty tanks and control residuals. Cold calcination of low and high NOx low activity waste (LAW) SRS Tank 44 and Hanford AN-107 simulants, respectively with metallic Al + Si powders was evaluated. It was found that a combination of Al and Si powders could be used as reducing agents to reduce the nitrate and nitrite content of both low and high NOx LAW to low enough levels to allow the LAW to be solidified directly by mixing it with metakaolin and allowing it to cure at 90 C. During room temperature reactions, NOx was reduced and nitrogen was emitted as N2 or NH3. This was an important finding because now one can pretreat LAW at ambient temperatures which provides a low-temperature alternative to thermal calcination. The significant advantage of using Al and Si metals for denitration/denitrition of the LAW is the fact that the supernate could potentially be treated in situ in the waste tanks themselves. Tank fill materials based upon a hydroceramic binder have been formulated from mixtures of metakaolinite, Class F fly ash and Class C flue gas desulphurization (FGD) ash mixed with various concentrations of NaOH solution. These harden over a period of hours or days depending on composition. A systematic study of properties of the tank fill materials (leachability) and ability to adsorb and hold residuals is under way.

Grutzeck, Michael

2005-06-01T23:59:59.000Z

229

Evolved Gas Analysis for High-alumina HLW (High Level Waste) Feed  

Science Conference Proceedings (OSTI)

Using the thermogravimetry coupled with gas chromatography-mass spectrometer, ... Tungstic Acid for Sorption of Uranium from Natural and Waste Waters and...

230

River Protection Project (RPP) Immobilized High Level Waste (HLW) Interim Storage Plan  

SciTech Connect

This document replaces HNF-1751, Revision 1. It incorporates updates to reflect changes in programmatic direction associated with the vitrification plant contract and associated DOE-ORP guidance. In addition it includes planning associated with failed/used melter and sample handling and disposition work scope. The document also includes format modifications and section numbering update consistent with CH2M HILL Hanford Group, Inc. procedures.

BRIGGS, M.G.

2000-09-22T23:59:59.000Z

231

Enhanced Sulfate Management in HLW Glass Formulations VSL12R2540-1 REV 0  

SciTech Connect

The Low Activity Waste (LAW) tanks that are scheduled to provide the Hanford Tank Waste Treatment and Immobilization Plant (WTP) with waste feeds contain significant amounts of sulfate. The sulfate content in the LAW feeds is sufficiently high that a separate molten sulfate salt phase may form on top of the glass melt during the vitrification process unless suitable glass formulations are employed and sulfate levels are controlled. Since the formation of the salt phase is undesirable from many perspectives, mitigation approaches had to be developed. Considerable progress has been made and reported by the Vitreous State Laboratory (VSL) in enhancing sulfate incorporation into LAW glass melts and developing strategies to manage and mitigate the risks associated with high-sulfate feeds.

Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Kot, Wing [The Catholic University of America, Washington, DC (United States); Gan, Hao [The Catholic University of America, Washington, DC (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States)

2012-11-13T23:59:59.000Z

232

Pyrochemical processing of Idaho Chemical Processing Plant (ICPP) High Level Waste (HLW) calcine  

SciTech Connect

Inertial force damping control by micromanipulator modulation is proposed to suppress the vibrations of a micro/macro-manipulator system. The proposed controller, developed using classical control theory, is added to the existing control system. The proposed controller uses real-time measurements of macro-manipulator flexibility to adjust the motion of the micro manipulator to counteract structural vibrations. Experimental studies using an existing micro/macro flexible link manipulator testbed demonstrate the effectiveness of the proposed approach to suppression of vibrations in the macro/micro-manipulator system using micromanipulator-based inertial active damping control.

Bronson, M.C.; Ebbinghaus, B.B.; Riley, D.C. [Lawrence Livermore National Lab., CA (United States); Nelson, L.; Del Debbio, J. [Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States)

1994-11-15T23:59:59.000Z

233

Management of nuclear materials and non-HLW | Department of Energy  

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

including plutonium, uranium, and nuclear waste in accordance with applicable statutes, DOE Orders and international commitments. Advice encompasses issues related to mixed oxide...

234

Report - Melter Testing of New High Bismuth HLW Formulations VSL-13R2770-1  

Science Conference Proceedings (OSTI)

The primary objective of the work described was to test two glasses formulated for a high bismuth waste stream on the DM100 melter system. Testing was designed to determine processing characteristics and production rates, assess the tendency for foaming, and confirm glass properties. The glass compositions tested were previously developed to maintain high waste loadings and processing rates while suppressing the foaming observed in previous tests

Kruger, Albert A.; Pegg, I. L.; Kot, W. K.; Gan, H.; Matlack, K. S.

2013-11-13T23:59:59.000Z

235

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

Science Conference Proceedings (OSTI)

To accomplish calcine treatment objectives, the Idaho Clean-up Project contractor, CWI, has chosen to immobilize the calcine in a glass-ceramic via the use of a Hot-Isostatic-Press (HIP); a treatment selection formally documented in a 2010 Record of Decision (ROD). Even though the HIP process may prove suitable for the calcine as specified in the ROD and validated in a number of past value engineering sessions, DOE is evaluating back-up treatment methods for the calcine as a result of the technical, schedule, and cost risk associated with the HIPing process. Consequently DOE HQ has requested DOE ID to make INL's bench-scale cold-crucible induction melter (CCIM) available for investigating its viability as a process alternate to calcine treatment. The waste form is the key component of immobilization of radioactive waste. Providing a solid, stable, and durable material that can be easily be stored is the rationale for immobilization of radioactive waste material in glass, ceramic, or glass-ceramics. Ceramic waste forms offer an alternative to traditional borosilicate glass waste forms. Ceramics can usually accommodate higher waste loadings than borosilicate glass, leading to smaller intermediate and long-term storage facilities. Many ceramic phases are known to possess superior chemical durability as compared to borosilicate glass. However, ceramics are generally multiphase systems containing many minor phase that make characterization and prediction of performance within a repository challenging. Additionally, the technologies employed in ceramic manufacture are typically more complex and expensive. Thus, many have proposed using glass-ceramics as compromise between in the more inexpensive, easier to characterize glass waste forms and the more durable ceramic waste forms. Glass-ceramics have several advantages over traditional borosilicate glasses as a waste form. Borosilicate glasses can inadvertently devitrify, leading to a less durable product that could crack during cooling and crystals may be prone to dissolution. By designing a glass-ceramics, the risks of deleterious effects from devitrification are removed. Furthermore, glass-ceramics have higher mechanical strength and impact strengths and possess greater chemical durability as noted above. Glass-ceramics should provide a waste form with the advantages of glass - ease of manufacture - with improved mechanical properties, thermal stability, and chemical durability. This report will cover aspects relevant for the validation of the CCIM use in the production of glass-ceramic waste forms.

James A. King; Vince Maio

2011-09-01T23:59:59.000Z

236

Corrosion mechanisms of low level vitrified radioactive waste in a loamy soil M.I. Ojovan1  

E-Print Network (OSTI)

Topic: Briefings by environmental groups, industry groups, pub- lic policy groups, and state, is the central authority responsi- ble for evaluating and supervising the nuclear industry's research and 1.95 meters in diameter. It is fabricated from forged steel with a stainless steel coating. The cask

Sheffield, University of

237

This is the title of the presentation on three lines if you need it  

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

41 41 In-Riser Ion Exchange: Resorcinol- Formaldehyde Maturation Bill King*, Dan McCabe, and Frank Pennebaker SRNL Environmental and Chemical Process Technology May 20, 2009 Office of Waste Processing Technical Exchange 2 SRNL-STI-2009-00341 Agenda Objectives Background Original Design Media Considered Current Conceptual Design Recent Progress Current Activities Project Goals SRNL - Environmental and Chemical Process Technology 3 SRNL-STI-2009-00341 Objectives SRNL - Environmental and Chemical Process Technology Develop in-tank system to decontaminate HLW salt solution at SRS and Hanford to accelerate salt processing and tank closure Vitrify Cs-137 or transfer to compliant tanks Support spherical RF maturation for various DOE applications Current concept: in-riser IX similar to Cesium

238

Quantity of Natural Gas Production Associated with Reported Wellhead Value  

U.S. Energy Information Administration (EIA) Indexed Site

Annual Download Series History Download Series History Definitions, Sources & Notes Definitions, Sources & Notes Show Data By: Data Series Area 2005 2006 2007 2008 2009 2010 View History U.S. 15,425,867 15,981,421 1980-2006 Alabama 285,237 274,176 259,062 246,747 225,666 212,769 1983-2010 Alaska 502,887 494,323 368,344 337,359 397,077 316,546 1983-2010 Arizona 211 588 634 503 695 165 1983-2010 Arkansas 190,533 193,491 269,886 446,551 680,613 936,600 1983-2010 California 274,817 278,933 268,016 263,107 241,916 251,559 1983-2010 Colorado 1,106,993 1,170,819 1,280,638 1,436,203 1,409,172 1,548,576 1983-2010 Florida NA NA NA NA NA NA 1983-2010 Illinois NA NA NA NA NA NA 1983-2010 Indiana

239

Quantity versus Quality in Off-Street Parking Requirements  

E-Print Network (OSTI)

off-street parking requirements does not restrict parking orrequirements if they are con- verted to residential uses. Los Angeles, for example, does

Mukhija, Vinit; Shoup, Donald

2006-01-01T23:59:59.000Z

240

Quality and Quantity Modeling of a Production Line  

E-Print Network (OSTI)

During the past three decades, the success of the Toyota Production System has spurred research in the area of manufacturing systems engineering. Two research fields, productivity and quality, have been extensively studied ...

Kim, Jongyoon

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

ANNUAL QUANTITY AND VALUE OF NATURAL GAS PRODUCTION REPORT FORM ...  

U.S. Energy Information Administration (EIA)

present in reservoir natural gas are water vapor, carbon dioxide, hydrogen sulfide, ... Shale Gas: Methane and other gases produced from wells that are open

242

ANNUAL QUANTITY AND VALUE OF NATURAL GAS PRODUCTION REPORT FORM ...  

U.S. Energy Information Administration (EIA)

reporting system within the limits imposed by the diversity of the data collection systems of the various producing States.

243

Table 7.6 Quantity of Purchased Energy Sources, 2010;  

U.S. Energy Information Administration (EIA) Indexed Site

Lime 103 1,502 * 1 3 * 4 * 9 327420 Gypsum 43 1,384 * * 37 * 0 0 * 327993 Mineral Wool 39 3,408 0 * 24 * 0 * * 331 Primary Metals 1,820 120,608 * 1 548 1 17 11 102 331111...

244

Examination of sharing fractions for prices and quantities  

Science Conference Proceedings (OSTI)

When the Household Model of Energy (HOME) and Commercial Sector Energy Model (CSEM) are run as modules in the Intermediate Future Forecasting System (IFFS), the interfacing variables (prices and consumption of fuels) have to be adjusted to their aggregated regional levels. Both HOME and CSEM operate at a level of 4 Census Regions whereas IFFS uses 10 federal Regions. This makes it necessary to aggregate the prices provided by IFFS to the 4 Census Regions and to disaggregate the sectoral consumption values calculated by HOME and CSEM to the 10 federal Regions. An examination of the historical fractions for consumption levels and prices by fuels and sectors (residential and commercial) was performed to substantiate the assumption that changes of these fractions over time are not significant. This assumption is presently employed in both HOME and CSEM. The fractions which are presently used were calculated for each fuel based on the consumption data for the year 1980. These fractions, once evaluated, are used for sharing both prices and consumption throughout the forecasting period.

Meyer, M.

1986-02-01T23:59:59.000Z

245

Immobilized High Level Waste (HLW) Interim Storage Alternative Generation and analysis and Decision Report 2nd Generation Implementing Architecture  

SciTech Connect

Two alternative approaches were previously identified to provide second-generation interim storage of Immobilized High-Level Waste (IHLW). One approach was retrofit modification of the Fuel and Materials Examination Facility (FMEF) to accommodate IHLW. The results of the evaluation of the FMEF as the second-generation IHLW interim storage facility and subsequent decision process are provided in this document.

CALMUS, R.B.

2000-09-14T23:59:59.000Z

246

Application of the Evacuated Canister System for Removing Residual Molten Glass From the West Valley Demonstration Project High-Level Waste Melter  

SciTech Connect

The principal mission of the West Valley Demonstration Project (WVDP) is to meet a series of objectives defined in the West Valley Demonstration Project Act (Public Law 96-368). Chief among these is the objective to solidify liquid high-level waste (HLW) at the WVDP site into a form suitable for disposal in a federal geologic repository. In 1982, the Secretary of Energy formally selected vitrification as the technology to be used to solidify HLW at the WVDP. One of the first steps in meeting the HLW solidification objective involved designing, constructing and operating the Vitrification (Vit) Facility, the WVDP facility that houses the systems and subsystems used to process HLW into stainless steel canisters of borosilicate waste-glass that satisfy waste acceptance criteria (WAC) for disposal in a federal geologic repository. HLW processing and canister production began in 1996. The final step in meeting the HLW solidification objective involved ending Vit system operations and shut ting down the Vit Facility. This was accomplished by conducting a discrete series of activities to remove as much residual material as practical from the primary process vessels, components, and associated piping used in HLW canister production before declaring a formal end to Vit system operations. Flushing was the primary method used to remove residual radioactive material from the vitrification system. The inventory of radioactivity contained within the entire primary processing system diminished by conducting the flushing activities. At the completion of flushing activities, the composition of residual molten material remaining in the melter (the primary system component used in glass production) consisted of a small quantity of radioactive material and large quantities of glass former materials needed to produce borosilicate waste-glass. A special system developed during the pre-operational and testing phase of Vit Facility operation, the Evacuated Canister System (ECS), was deployed at the West Valley Demonstration Project to remove this radioactively dilute, residual molten material from the melter before Vit system operations were brought to a formal end. The ECS consists of a stainless steel canister of the same size and dimensions as a standard HLW canister that is equipped with a special L-shaped snorkel assembly made of 304L stainless steel. Both the canister and snorkel assembly fit into a stainless steel cage that allows the entire canister assembly to be positioned over the melter as molten glass is drawn out by a vacuum applied to the canister. This paper describes the process used to prepare and apply the ECS to complete molten glass removal before declaring a formal end to Vit system operations and placing the Vit Facility into a safe standby mode awaiting potential deactivation.

May, Joseph J.; Dombrowski, David J.; Valenti, Paul J.; Houston, Helene M.

2003-02-27T23:59:59.000Z

247

HIGH LEVEL WASTE SLUDGE BATCH 4 VARIABILITY STUDY  

Science Conference Proceedings (OSTI)

The Defense Waste Processing Facility (DWPF) is preparing for vitrification of High Level Waste (HLW) Sludge Batch 4 (SB4) in early FY2007. To support this process, the Savannah River National Laboratory (SRNL) has provided a recommendation to utilize Frit 503 for vitrifying this sludge batch, based on the composition projection provided by the Liquid Waste Organization on June 22, 2006. Frit 418 was also recommended for possible use during the transition from SB3 to SB4. A critical step in the SB4 qualification process is to demonstrate the applicability of the durability models, which are used as part of the DWPF's process control strategy, to the glass system of interest via a variability study. A variability study is an experimentally-driven assessment of the predictability and acceptability of the quality of the vitrified waste product that is anticipated from the processing of a sludge batch. At the DWPF, the durability of the vitrified waste product is not directly measured. Instead, the durability is predicted using a set of models that relate the Product Consistency Test (PCT) response of a glass to the chemical composition of that glass. In addition, a glass sample is taken during the processing of that sludge batch, the sample is transmitted to SRNL, and the durability is measured to confirm acceptance. The objective of a variability study is to demonstrate that these models are applicable to the glass composition region anticipated during the processing of the sludge batch - in this case the Frit 503 - SB4 compositional region. The success of this demonstration allows the DWPF to confidently rely on the predictions of the durability/composition models as they are used in the control of the DWPF process.

Fox, K; Tommy Edwards, T; David Peeler, D; David Best, D; Irene Reamer, I; Phyllis Workman, P

2006-10-02T23:59:59.000Z

248

Crystal Reports - sum4.rpt  

Office of Environmental Management (EM)

Waste/Contaminated Media Waste/Contaminated Media and SNF Inventories By Program (Sum-4) Current Year: 2000 Pr o g r am HL W HL W -V i trifie d * T RU M L L W L L W O T HER ** S N F *** Qu a n t ity (m 3) Qu an tit y (NC ) Qu an tity (m 3) Qu a n tity (m 3 ) Qu an tit y (m 3) Qu an tity (m 3) Qu an tit y (M T HM ) Office of Defense Programs 0.00 0.00 628.20 391.82 1,843.59 0.00 2.44 Office of Environmental Management 353,500.78 1,201.00 110,447.25 45,869.38 156,965.74 3,946.00 2,442.12 Nuclear Energy 0.00 0.00 60.90 226.20 0.00 0.00 22.10 Non-DOE sources 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Office of Science 0.00 0.00 100.29 60.50 100.25 0.00 0.61 *Vitrified HLW quantites are reported in # of HLW Canisters. **Other includes "Unspecified" and 11(e)2 waste types.

249

Chemical Composiiton Analysis of INEEL Phase 3 Glasses: Task Technical and QA Plan  

SciTech Connect

For about four decades radioactive wastes have been collected and calcined from nuclear fuels reprocessing at the Idaho Chemical Processing Plant (ICPP). Over this time span, secondary radioactive waste from decontamination, laboratory activities and fuels storage activities have also been collected and stored as liquid. These liquid high-activity wastes (HAW) are collectively called Sodium Bearing Wastes (SBW). Currently about 5.7 million liters of these wastes are temporarily stored in stainless steel tanks at the Idaho National Engineering and Environmental Laboratory (INEEL). Vitrification is being considered as a treatment option for SBW. The resulting glass can be sent to either the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, as remote handled transuranic waste (RH-TRU) or to the federal geologic repository for final disposal. In addition to the SBW, roughly 4,000 m3 of calcined high-level wastes (HLW) are currently being stored at INEEL in stainless steel bin sets. These calcined HLW may also be vitrified, either with or without a dissolution and separation process, and sent to the federal geologic repository for final disposal.

Peeler, D.

2000-08-11T23:59:59.000Z

250

Separation of strontium-90 from Hanford high-level radioactive waste  

SciTech Connect

Current guidelines for disposing of high-level radioactive wastes stored in underground tanks at the US Department of Energy`s Hanford Site call for vitrifying high-level waste (HLW) in borosilicate glass and disposing the glass canisters in a deep geologic repository. Disposition of the low-level waste (LLW) is yet to be determined, but it will likely be immobilized in a glass matrix and disposed of on site. To lower the radiological risk associated with the LLW form, methods are being developed to separate {sup 90}Sr from the bulk waste material so this isotope can be routed to the HLW stream. A solvent extraction method is being investigated to separate {sup 90}Sr from acid-dissolved Hanford tank wastes. Results of experiments with actual tank waste indicate that this method can be used to achieve separation of {sup 90}Sr from the bulk waste components. Greater than 99% of the {sup 90}Sr was removed from an acidic dissolved sludge solution by extraction with di-tbutylcyclohexano-18-crown-6 in 1-octanol (the SREX process). The major sludge components were not extracted.

Lumetta, G.J.; Wagner, M.J.; Jones, E.O.

1993-10-01T23:59:59.000Z

251

Vitrification of low-level and mixed wastes  

SciTech Connect

The US Department of Energy (DOE) and nuclear utilities have large quantities of low-level and mixed wastes that must be treated to meet repository performance requirements, which are likely to become even more stringent. The DOE is developing cost-effective vitrification methods for producing durable waste forms. However, vitrification processes for high-level wastes are not applicable to commercial low-level wastes containing large quantities of metals and small amounts of fluxes. New vitrified waste formulations are needed that are durable when buried in surface repositories.

Johnson, T.R.; Bates, J.K.; Feng, Xiangdong

1994-12-31T23:59:59.000Z

252

Generalized Test Plan for the Vitrification of Simulated High-Level -Waste Calcine in the Idaho National Laboratorys Bench -Scale Cold Crucible Induction Melter  

SciTech Connect

This Preliminary Idaho National Laboratory (INL) Test Plan outlines the chronological steps required to initially evaluate the validity of vitrifying INL surrogate (cold) High-Level-Waste (HLW) solid particulate calcine in INL's Cold Crucible Induction Melter (CCIM). Its documentation and publication satisfies interim milestone WP-413-INL-01 of the DOE-EM (via the Office of River Protection) sponsored work package, WP 4.1.3, entitled 'Improved Vitrification' The primary goal of the proposed CCIM testing is to initiate efforts to identify an efficient and effective back-up and risk adverse technology for treating the actual HLW calcine stored at the INL. The calcine's treatment must be completed by 2035 as dictated by a State of Idaho Consent Order. A final report on this surrogate/calcine test in the CCIM will be issued in May 2012-pending next fiscal year funding In particular the plan provides; (1) distinct test objectives, (2) a description of the purpose and scope of planned university contracted pre-screening tests required to optimize the CCIM glass/surrogate calcine formulation, (3) a listing of necessary CCIM equipment modifications and corresponding work control document changes necessary to feed a solid particulate to the CCIM, (4) a description of the class of calcine that will be represented by the surrogate, and (5) a tentative tabulation of the anticipated CCIM testing conditions, testing parameters, sampling requirements and analytical tests. Key FY -11 milestones associated with this CCIM testing effort are also provided. The CCIM test run is scheduled to be conducted in February of 2012 and will involve testing with a surrogate HLW calcine representative of only 13% of the 4,000 m3 of 'hot' calcine residing in 6 INL Bin Sets. The remaining classes of calcine will have to be eventually tested in the CCIM if an operational scale CCIM is to be a feasible option for the actual INL HLW calcine. This remaining calcine's make-up is HLW containing relatively high concentrations of zirconium and aluminum, representative of the cladding material of the reprocessed fuel that generated the calcine. A separate study to define the CCIM testing needs of these other calcine classifications in currently being prepared under a separate work package (WP-0) and will be provided as a milestone report at the end of this fiscal year.

Vince Maio

2011-08-01T23:59:59.000Z

253

HIGH-LEVEL WASTE FEED CERTIFICATION IN HANFORD DOUBLE-SHELL TANKS  

SciTech Connect

The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. DOE's River Protection Project (RPP) mission modeling and WTP facility modeling assume that individual 3785 cubic meter (l million gallon) HLW feed tanks are homogenously mixed, representatively sampled, and consistently delivered to the WTP. It has been demonstrated that homogenous mixing ofHLW sludge in Hanford DSTs is not likely achievable with the baseline design thereby causing representative sampling and consistent feed delivery to be more difficult. Inconsistent feed to the WTP could cause additional batch-to-batch operational adjustments that reduce operating efficiency and have the potential to increase the overall mission length. The Hanford mixing and sampling demonstration program will identify DST mixing performance capability, will evaluate representative sampling techniques, and will estimate feed batch consistency. An evaluation of demonstration program results will identify potential mission improvement considerations that will help ensure successful mission completion. This paper will discuss the history, progress, and future activities that will define and mitigate the mission risk.

THIEN MG; WELLS BE; ADAMSON DJ

2010-01-14T23:59:59.000Z

254

High Level Waste Feed Certification in Hanford Double Shell Tanks  

SciTech Connect

The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. DOEs River Protection Project (RPP) mission modeling and WTP facility modeling assume that individual 3785 cubic meter (1 million gallon) HLW feed tanks are homogenously mixed, representatively sampled, and consistently delivered to the WTP. It has been demonstrated that homogenous mixing of HLW sludge in Hanford DSTs is not likely achievable with the baseline design thereby causing representative sampling and consistent feed delivery to be more difficult. Inconsistent feed to the WTP could cause additional batch to batch operational adjustments that reduces operating efficiency and has the potential to increase the overall mission length. The Hanford mixing and sampling demonstration program will identify DST mixing performance capability, will evaluate representative sampling techniques, and will estimate feed batch consistency. An evaluation of demonstration program results will identify potential mission improvement considerations that will help ensure successful mission completion. This paper will discuss the history, progress, and future activities that will define and mitigate the mission risk.

Thien, Micheal G.; Wells, Beric E.; Adamson, Duane J.

2010-03-01T23:59:59.000Z

255

The effect of vitrification technology on waste loading  

SciTech Connect

Radioactive wastes on the Hanford Site are going to be permanently disposed of by incorporation into a durable glass. These wastes will be separated into low and high-level portions, and then vitrified. The low-level waste (LLW) is water soluble. Its vitrifiable part (other than off-gas) contains approximately 80 wt% Na{sub 2}O, the rest being Al{sub 2}O{sub 3}, P{sub 2}O{sub 5}, K{sub 2}O, and minor components. The challenge is to formulate durable LLW glasses with as high Na{sub 2}O content as possible by optimizing the additions of SiO{sub 2}, Al{sub 2}O{sub 3}, B{sub 2}O{sub 3}, CaO, and ZrO{sub 2}. This task will not be simple, considering the non-linear and interactive nature of glass properties as a function of composition. Once developed, the LLW glass, being similar in composition to commercial glasses, is unlikely to cause major processing problems, such as crystallization or molten salt segregation. For example, inexpensive LLW glass can be produced in a high-capacity Joule-heated melter with a cold cap to minimize volatilization. The high-level waste (HLW) consists of water-insoluble sludge (Fe{sub 2}O{sub 3}, Al{sub 2}O{sub 3}, ZrO{sub 2}, Cr{sub 2}O{sub 3}, NiO, and others) and a substantial water-soluble residue (Na{sub 2}O). Most of the water-insoluble components are refractory; i.e., their melting points are above the glass melting temperature. With regard to product acceptability, the maximum loading of Hanford HLW in the glass is limited by product durability, not by radiolytic heat generation. However, this maximum may not be achievable because of technological constraints imposed by melter feed rheology, frit properties, and glass melter limits. These restrictions are discussed in this paper. 38 refs.

Hrma, P.R.; Smith, P.A.

1994-08-01T23:59:59.000Z

256

NEXT GENERATION MELTER(S) FOR VITRIFICATION OF HANFORD WASTE STATUS AND DIRECTION  

SciTech Connect

Vitrification technology has been selected to treat high-level waste (HLW) at the Hanford Site, the West Valley Demonstration Project and the Savannah River Site (SRS), and low activity waste (LAW) at Hanford. In addition, it may potentially be applied to other defense waste streams such as sodium bearing tank waste or calcine. Joule-heated melters (already in service at SRS) will initially be used at the Hanford Site's Waste Treatment and Immobilization Plant (WTP) to vitrify tank waste fractions. The glass waste content and melt/production rates at WTP are limited by the current melter technology. Significant reductions in glass volumes and mission life are only possible with advancements in melter technology coupled with new glass formulations. The Next Generation Melter (NGM) program has been established by the U.S. Department of Energy's (DOE's), Environmental Management Office of Waste Processing (EM-31) to develop melters with greater production capacity (absolute glass throughput rate) and the ability to process melts with higher waste fractions. Advanced systems based on Joule-Heated Ceramic Melter (JHCM) and Cold Crucible Induction Melter (CCIM) technologies will be evaluated for HLW and LAW processing. Washington River Protection Solutions (WRPS), DOE's tank waste contractor, is developing and evaluating these systems in cooperation with EM-31, national and university laboratories, and corporate partners. A primary NGM program goal is to develop the systems (and associated flowsheets) to Technology Readiness Level 6 by 2016. Design and testing are being performed to optimize waste glass process envelopes with melter and balance of plant requirements. A structured decision analysis program will be utilized to assess the performance of the competing melter technologies. Criteria selected for the decision analysis program will include physical process operations, melter performance, system compatibility and other parameters.

RAMSEY WG; GRAY MF; CALMUS RB; EDGE JA; GARRETT BG

2011-01-13T23:59:59.000Z

257

Depleted uranium: A DOE management guide  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy (DOE) has a management challenge and financial liability in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. The annual storage and maintenance cost is approximately $10 million. This report summarizes several studies undertaken by the DOE Office of Technology Development (OTD) to evaluate options for long-term depleted uranium management. Based on studies conducted to date, the most likely use of the depleted uranium is for shielding of spent nuclear fuel (SNF) or vitrified high-level waste (HLW) containers. The alternative to finding a use for the depleted uranium is disposal as a radioactive waste. Estimated disposal costs, utilizing existing technologies, range between $3.8 and $11.3 billion, depending on factors such as applicability of the Resource Conservation and Recovery Act (RCRA) and the location of the disposal site. The cost of recycling the depleted uranium in a concrete based shielding in SNF/HLW containers, although substantial, is comparable to or less than the cost of disposal. Consequently, the case can be made that if DOE invests in developing depleted uranium shielded containers instead of disposal, a long-term solution to the UF{sub 6} problem is attained at comparable or lower cost than disposal as a waste. Two concepts for depleted uranium storage casks were considered in these studies. The first is based on standard fabrication concepts previously developed for depleted uranium metal. The second converts the UF{sub 6} to an oxide aggregate that is used in concrete to make dry storage casks.

NONE

1995-10-01T23:59:59.000Z

258

FINAL REPORT DURAMELTER 100 HLW SIMULANT VALIDATION TESTS WITH C-106/AY-102 FEEDS VSL-05R5710-1 REV 0 6/2/05  

Science Conference Proceedings (OSTI)

The principal objectives of the DM100 tests were to determine the processing characteristics of several C-106/AY102 feeds derived from simulants prepared by different methods, which result in different physical characteristics of the feed. The VSL simulant used in a previous test was prepared by the direct hydroxide method, which was the method used for feed preparation in the bulk of previous VSL melter testing. The NOAH Technologies Corporation modified-rheology simulant was prepared to the same composition as the VSL simulant using a method that resulted in rheological properties closer to those of certain actual waste samples. The SIPP simulant was produced by processing a co-precipitated waste simulant through a non-radioactive pilot scale semi-integrated pretreatment facility. The general intent of these tests was to provide a basis for determining whether the variations in rheology or other feed physical characteristics arising from the different methods of simulant preparation have significant effects on the processing characteristics of the feed in the melter. Completion of the test objectives is detailed in a table.

KRUGER AA; MATLACK KS; GONG W; PEGG IL

2011-12-29T23:59:59.000Z

259

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

SciTech Connect

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

Grutzeck, Michael W.

2005-06-27T23:59:59.000Z

260

Comments Regarding In-Drift Chemistry Related to Corrosion of Containment Barriers at the Candidate Spent Fuel and HLW Repository at Yucca Mountain, Nevada  

Science Conference Proceedings (OSTI)

The Nuclear Waste Technical Review Board (NWTRB) has postulated a scenario for the formation of a deliquescent divalent cation-chloride brine that they believe could lead to earlier than expected penetration of nuclear waste packages at a repository at Yucca Mountain, and thereby compromise compliance with established regulatory criteria. The present report documents the results of an independent technical analysis of this scenario carried out on behalf of EPRI. The analysis has specifically examined and...

2004-05-31T23:59:59.000Z

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261

ACTUAL-WASTE TESTS OF ENHANCED CHEMICAL CLEANING FOR RETRIEVAL OF SRS HLW SLUDGE TANK HEELS AND DECOMPOSITION OF OXALIC ACID  

Science Conference Proceedings (OSTI)

Savannah River National Laboratory conducted a series of tests on the Enhanced Chemical Cleaning (ECC) process using actual Savannah River Site waste material from Tanks 5F and 12H. Testing involved sludge dissolution with 2 wt% oxalic acid, the decomposition of the oxalates by ozonolysis (with and without the aid of ultraviolet light), the evaporation of water from the product, and tracking the concentrations of key components throughout the process. During ECC actual waste testing, the process was successful in decomposing oxalate to below the target levels without causing substantial physical or chemical changes in the product sludge.

Martino, C.; King, W.; Ketusky, E.

2012-01-12T23:59:59.000Z

262

Washing and caustic leaching of Hanford tank sludge: Results of FY 1997 studies  

Science Conference Proceedings (OSTI)

The current plan for remediating the Hanford tank farms consists of waste retrieval, pretreatment, treatment (immobilization), and disposal. The tank wastes will be partitioned into high-level and low-level fractions. The HLW will be immobilized in a borosilicate glass matrix; the resulting glass canisters will then be disposed of in a geologic repository. Because of the expected high cost of HLW vitrification and geologic disposal, pretreatment processes will be implemented to reduce the volume of immobilized high-level waste (IHLW). Caustic leaching (sometimes referred to as enhanced sludge washing or ESW) represents the baseline method for pretreating Hanford tank sludges. Caustic leaching is expected to remove a large fraction of the Al, which is present in large quantities in Hanford tank sludges. A significant portion of the P is also expected to be removed from the sludge by metathesis of water-insoluble metal phosphates to insoluble hydroxides and soluble Na{sub 3}PO{sub 4}. Similar metathesis reactions can occur for insoluble sulfate salts, allowing the removal of sulfate from the HLW stream. This report describes the sludge washing and caustic leaching tests performed at the Pacific Northwest National Laboratory in FY 1996. The sludges used in this study were taken from Hanford tanks AN-104, BY-108, S-101, and S-111.

Lumetta, G.J.; Burgeson, I.E.; Wagner, M.J.; Liu, J.; Chen, Y.L.

1997-08-01T23:59:59.000Z

263

Hydrogeology, chemical and microbial activity measurement through deep permafrost  

E-Print Network (OSTI)

Ethene % Ethane % Propene % Propane % Butene % iso-Butane %Butane C1/(C2 + C3) HLW-03-28, Purge 4 ND HLW-03-28, Purge 6

Stotler, R.L.

2010-01-01T23:59:59.000Z

264

Office of the Assistant General Counsel for Civilian Nuclear...  

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

Waste (HLW) and Spent Nuclear Fuel (SNF) Management of Nuclear Materials and Non-HLW Nuclear Fuel Cycle Energy Research and Development Non-Proliferation Nuclear Regulatory...

265

Clean option: An alternative strategy for Hanford Tank Waste Remediation. Volume 2, Detailed description of first example flowsheet  

SciTech Connect

Disposal of high-level tank wastes at the Hanford Site is currently envisioned to divide the waste between two principal waste forms: glass for the high-level waste (HLW) and grout for the low-level waste (LLW). The draft flow diagram shown in Figure 1.1 was developed as part of the current planning process for the Tank Waste Remediation System (TWRS), which is evaluating options for tank cleanup. The TWRS has been established by the US Department of Energy (DOE) to safely manage the Hanford tank wastes. It includes tank safety and waste disposal issues, as well as the waste pretreatment and waste minimization issues that are involved in the ``clean option`` discussed in this report. This report describes the results of a study led by Pacific Northwest Laboratory to determine if a more aggressive separations scheme could be devised which could mitigate concerns over the quantity of the HLW and the toxicity of the LLW produced by the reference system. This aggressive scheme, which would meet NRC Class A restrictions (10 CFR 61), would fit within the overall concept depicted in Figure 1.1; it would perform additional and/or modified operations in the areas identified as interim storage, pretreatment, and LLW concentration. Additional benefits of this scheme might result from using HLW and LLW disposal forms other than glass and grout, but such departures from the reference case are not included at this time. The evaluation of this aggressive separations scheme addressed institutional issues such as: radioactivity remaining in the Hanford Site LLW grout, volume of HLW glass that must be shipped offsite, and disposition of appropriate waste constituents to nonwaste forms.

Swanson, J.L.

1993-09-01T23:59:59.000Z

266

MRS/IS facility co-located with a repository: preconceptual design and life-cycle cost estimates  

SciTech Connect

A program is described to examine the various alternatives for monitored retrievable storage (MRS) and interim storage (IS) of spent nuclear fuel, solidified high-level waste (HLW), and transuranic (TRU) waste until appropriate geologic repository/repositories are available. The objectives of this study are: (1) to develop a preconceptual design for an MRS/IS facility that would become the principal surface facility for a deep geologic repository when the repository is opened, (2) to examine various issues such as transportation of wastes, licensing of the facility, and environmental concerns associated with operation of such a facility, and (3) to estimate the life cycle costs of the facility when operated in response to a set of scenarios which define the quantities and types of waste requiring storage in specific time periods, which generally span the years from 1990 until 2016. The life cycle costs estimated in this study include: the capital expenditures for structures, casks and/or drywells, storage areas and pads, and transfer equipment; the cost of staff labor, supplies, and services; and the incremental cost of transporting the waste materials from the site of origin to the MRS/IS facility. Three scenarios are examined to develop estimates of life cycle costs of the MRS/IS facility. In the first scenario, HLW canisters are stored, starting in 1990, until the co-located repository is opened in the year 1998. Additional reprocessing plants and repositories are placed in service at various intervals. In the second scenario, spent fuel is stored, starting in 1990, because the reprocessing plants are delayed in starting operations by 10 years, but no HLW is stored because the repositories open on schedule. In the third scenario, HLW is stored, starting in 1990, because the repositories are delayed 10 years, but the reprocessing plants open on schedule.

Smith, R.I.; Nesbitt, J.F.

1982-11-01T23:59:59.000Z

267

Preliminary ILAW Formulation Algorithm Description, 24590 LAW RPT-RT-04-0003, Rev. 1  

SciTech Connect

The U.S. Department of Energy (DOE), Office of River Protection (ORP), has contracted with Bechtel National, Inc. (BNI) to design, construct, and commission the Hanford Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site (DOE 2000). This plant is designed to operate for 40 years and treat roughly 50 million gallons of mixed hazardous high-level waste (HLW) stored in 177 underground tanks at the Hanford Site. The process involves separating the hight-level and low-activity waste (LAW) fractions through filtration, leaching, Cs ion exchange, and precipitation. Each fraction will be separately vitrified into borosilicate waste glass. This report documents the initial algorithm for use by Hanford WTP in batching LAW and glass-forming chemicals (GFCs) in the LAW melter feed preparation vessel (MFPV). Algorithm inputs include the chemical analyses of the pretreated LAW in the concentrate receipt vessel (CRV), the volume of the MFPV heel, and the compositions of individual GFCs. In addition to these inputs, uncertainties in the LAW composition and processing parameters are included in the algorithm.

Kruger, Albert A.; Kim, Dong-Sang; Vienna, John D.

2013-12-03T23:59:59.000Z

268

Evaluation of Flygt Propeller Xixers for Double Shell Tank (DST) High Level Waste Auxiliary Solids Mobilization  

Science Conference Proceedings (OSTI)

The River Protection Project (RPP) is planning to retrieve radioactive waste from the single-shell tanks (SST) and double-shell tanks (DST) underground at the Hanford Site. This waste will then be transferred to a waste treatment plant to be immobilized (vitrified) in a stable glass form. Over the years, the waste solids in many of the tanks have settled to form a layer of sludge at the bottom. The thickness of the sludge layer varies from tank to tank, from no sludge or a few inches of sludge to about 15 ft of sludge. The purpose of this technology and engineering case study is to evaluate the Flygt{trademark} submersible propeller mixer as a potential technology for auxiliary mobilization of DST HLW solids. Considering the usage and development to date by other sites in the development of this technology, this study also has the objective of expanding the knowledge base of the Flygt{trademark} mixer concept with the broader perspective of Hanford Site tank waste retrieval. More specifically, the objectives of this study delineated from the work plan are described.

PACQUET, E.A.

2000-07-20T23:59:59.000Z

269

Trade-offs in size, quantity and reliability of generalized nuclear power plants: a preliminary assessment  

SciTech Connect

An approximate method is used to estimate the effects of system reliability on optimal nuclear plant size, taking into account also scale factors and manufacturing learning curve slopes. The method is used to estimate the additional effective capability gained by adding units of different sizes to an existing electrical system. The number of additional units proves to be sensitive to forced outrage rate, estimated here from trends in US light-water reactors from 1971 to 1980. The relative cost of added units ranging in size from 200 to 800 MW is determined as a function of the parameters: scale factor and learning curve slope. The results generally corrobate the trends found in an earlier study in which the effect of reliability on required installed capacity was not explicitly considered. Optimal plant size decreases with weaker scale effects and stronger learning curve effects. Reliability considerations further reduce the optimal plant size, but the relative reduction is apparently not as great with steeper learning curves. This is a plausible finding inasmuch as the reduction in numbers of additional units due to reliability considerations will affect cost most where the learning curve is steepest. 9 refs., 4 figs., 3 tabs.

Hill, D.

1985-04-01T23:59:59.000Z

270

Measurement techniques for local and global fluid dynamic quantities in two and three phase systems  

Science Conference Proceedings (OSTI)

Available measurement techniques for evaluation of global and local phase holdups, instantaneous and average phase velocities and for the determination of bubble sizes in gas-liquid and gas-liquid-solid systems are reviewed. Advantages and disadvantages of various techniques are discussed. Particular emphasis is placed on identifying methods that can be employed on large scale, thick wall, high pressure and high temperature reactors used in the manufacture of fuels and chemicals from synthesis gas and its derivatives.

Kumar, S.; Dudukovic, M.P.; Toseland, B.A.

1998-01-01T23:59:59.000Z

271

Measurement techniques for local and global fluid dynamic quantities in two and three phase systems  

SciTech Connect

This report presents a critical review of the methods available for assessing the fluid dynamic parameters in large industrial two and three phase bubble column and slurry bubble column reactors operated at high pressure and temperature. The physical principles behind various methods are explained, and the basic design of the instrumentation needed to implement each measurement principle is discussed. Fluid dynamic properties of interest are: gas, liquid and solids holdup and their axial and radial distribution as well as the velocity distribution of the two (bubble column) or three phases (slurry bubble column). This information on operating pilot plant and plant reactors is essential to verify the computational fluid dynamic codes as well as scale-up rules used in reactor design. Without such information extensive and costly scale-up to large reactors that exploit syngas chemistries, and other reactors in production of fuels and chemicals, cannot be avoided. In this report, available measurement techniques for evaluation of global and local phase holdups, instantaneous and average phase velocities and for the determination of bubble sizes in gas-liquid and gas-liquid-solid systems are reviewed. Advantages and disadvantages of various techniques are discussed. Particular emphasis is placed on identifying methods that can be employed on large scale, thick wall, high pressure and high temperature reactors used in the manufacture of fuels and chemicals from synthesis gas and its derivatives.

Kumar, S.; Dudukovic, M.P. [Washington Univ., St. Louis, MO (United States). Chemical Reaction Engineering Lab.; Toseland, B.A. [Air Products and Chemicals, Inc., Lehigh Valley, PA (United States)

1996-03-01T23:59:59.000Z

272

ITEM NO. SUPPLIES/SERVICES QUANTITY UNIT UNIT PRICE AMOUNT NAME...  

National Nuclear Security Administration (NNSA)

POLYCHLORINATED BIPHENYL FEDERAL FACILITIES COMPLIANCE AGREEMENT EP PUBLIC LAW 104-113 NATIONAL TECHNOLOGY TRANSFER AND ADVANCEMENT ACT OF 1995 EG, CP SECTION J, ATTACHMENT E...

273

Monitoring North Pacific Heat Content Variability: An Indicator of Fish Quantity?  

Science Conference Proceedings (OSTI)

Fields of modeled sea surface heights and temperatures are used to develop an algorithm to monitor the low-frequency heat content variability of the North Pacific's midlatitudes associated with regime shifts in the circulation patterns of the ...

R. Tokmakian

2003-11-01T23:59:59.000Z

274

ITEM NO. SUPPLIES/SERVICES QUANTITY UNIT UNIT PRICE AMOUNT NAME OF OFFEROR OR CONTRACTOR  

E-Print Network (OSTI)

Not Pass 2 Universal Waste Systems, Inc. CNG Refueling Station for Refuse trucks with Public Access $200 Infrastructure Project $470,600 $470,600 89.7% Awardee 19 SCAQMD Ontario 76 CNG Infrastructure Installation $300 Alternative Fuel CNG Station $195,600 $195,600 81.9% Awardee 18 South Coast Air Quality Management District

275

A Probabilistic Graphical Approach to Computing Electricity Price Duration Curves under Price and Quantity Competition.  

Science Conference Proceedings (OSTI)

The electricity price duration curve (EPDC) represents the probability distribution function of the electricity price considered as a random variable. The price uncertainty comes both from the demand side and the supply side, since the load varies continuously, ...

Pascal Michaillat; Shmuel Oren

2007-01-01T23:59:59.000Z

276

(2) Quantities and Prices of Animal Manure and Gaseous Fuels Generated:  

E-Print Network (OSTI)

In this context, we are defining animal manure as the excrement of livestock reared in agricultural operations as well as straw, sawdust, and other residues used as animal bedding. Gaseous fuels may be derived from municipal and industrial landfills (landfill gas) or from animal manure and solid biomass such as crop silage or the organic fraction of MSW (biogas). Both landfill gas and biogas are generated via anaerobic digestion, a multi-stage process whereby bacteria convert carbohydrates, fats, and proteins to methane (Evans 2001). EPA does not consider these materials to be wastes in themselves, when used as fuel, but rather materials derived from wastes.

unknown authors

2010-01-01T23:59:59.000Z

277

Effect of seasonal changes in quantities of biowaste on full scale anaerobic digester performance  

Science Conference Proceedings (OSTI)

A 750,000 l digester located in Roppen/Austria was studied over a 2-year period. The concentrations and amounts of CH{sub 4}, H{sub 2}, CO{sub 2} and H{sub 2}S and several other process parameters like temperature, retention time, dry weight and input of substrate were registered continuously. On a weekly scale the pH and the concentrations of NH{sub 4}{sup +}-N and volatile fatty acids (acetic, butyric, iso-butyric, propionic, valeric and iso-valeric acid) were measured. The data show a similar pattern of seasonal gas production over 2 years of monitoring. The consumption of VFA and not the hydrogenotrophic CH{sub 4} production appeared to be the limiting factor for the investigated digestion process. Whereas the changes in pH and the concentrations of most VFA did not correspond with changes in biogas production, the ratio of acetic to propionic acid and the concentration of H{sub 2} appeared to be useful indicators for reactor performance. However, the most influential factors for the anaerobic digestion process were the amount and the quality of input material, which distinctly changed throughout the year.

Illmer, P. [University of Innsbruck, Institute of Microbiology, Technikerstr. 25, A-6020 Innsbruck (Austria)], E-mail: Paul.Illmer@uibk.ac.at; Gstraunthaler, G. [Abfallbeseitigungsverband Westtirol, Breite Mure, A-6426 Roppen (Austria)

2009-01-15T23:59:59.000Z

278

Comparison of Variations in Atmospheric Quantities with Sea Surface Temperature Variations in the Equatorial Eastern Pacific  

Science Conference Proceedings (OSTI)

Sea surface temperature (SST) variations in the equatorial eastern Pacific (010S, 18090W) are compared with variations in atmospheric temperature, circulation, rainfall and trace-constituent amount. Significant at the 99.9% level (taking into ...

J. K. Angell

1981-02-01T23:59:59.000Z

279

Method and apparatus for dispensing small quantities of mercury from evacuated and sealed glass capsules  

DOE Patents (OSTI)

A technique is disclosed for opening an evacuated and sealed glass capsule containing a material that is to be dispensed which has a relatively high vapor pressure such as mercury. The capsule is typically disposed in a discharge tube envelope. The technique involves the use of a first light source imaged along the capsule and a second light source imaged across the capsule substantially transversely to the imaging of the first light source. Means are provided for constraining a segment of the capsule along its length with the constraining means being positioned to correspond with the imaging of the second light source. These light sources are preferably incandescent projection lamps. The constraining means is preferably a multiple looped wire support. 6 figs.

Grossman, M.W.; George, W.A.; Pai, R.Y.

1985-08-13T23:59:59.000Z

280

Methods to prepare large quantities of Mesoporous TiO2 B ...  

High power and high energy density are essential to batteries for applications in electric vehicles, stationary energy ... safety and rate capability.

Note: This page contains sample records for the topic "vitrified hlw quantities" 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.


281

Improper Rotations'Symmetry of Electromagnetic Field and New Conserved Quantity of Complex Fields  

E-Print Network (OSTI)

It is established the partition of linear space $$ over the field of genuine scalars and pseudoscalars, the vectors in which are sets of contravariant and covariant electromagnetic field tensors and pseudotensors $F^{\\mu\

Yearchuck, D

2009-01-01T23:59:59.000Z

282

Environmental Protection Agency 61.220 (iii) The quantity (in pounds) of phos-  

E-Print Network (OSTI)

for the phosphogypsum; (vi) A copy of each certification doc- ument which accompanied the phos- phogypsum at the time or transfer of phosphogypsum to a person other than an agricultural end- user, the distributor, retailer information: (1) The name and address of the per- son in charge of the activity involving use of phosphogypsum

283

Modeling Physical Quantities in Industrial Systems using Fluid Stochastic Petri Nets  

E-Print Network (OSTI)

and production systems are governed by discrete state digital controllers whose internal state transitions. When a continuous variable reaches a boundary, it triggers a state change. In this way, the transition discrete places, representing four system states, three timed transition and two fluid places which

Gribaudo, Marco

284

The Influence of the Mushroom Compost Application on the Microorganism Quantity of Reclamated Soil  

Science Conference Proceedings (OSTI)

The mushroom compost which was produced from farm can be used to improving the quality of the reclamated soil. On the one hand, the question about environmental pollution made by the mushroom compost is solved, and on the other hand, it can improve the ... Keywords: mushroom material, micro-organisms, soil quality

Liu Xueran; Li Xinju; Li Bing

2010-03-01T23:59:59.000Z

285

Greenhouse Warming and Efficient Climate Protection Policy, with discussion of Regulation by Price or by Quantity  

E-Print Network (OSTI)

are a domestic coal-fired plant costing $800 per kilowatt,gigawatt plant nuclear or renewable rather than coal-fired.plant in China nuclear/renewable rather than coal fired, all

Lydon, Peter

2002-01-01T23:59:59.000Z

286

Analysis of statistical quantities in simulation of fluidized beds Kengo Ichiki* and Hisao Hayakawa  

E-Print Network (OSTI)

categories: two-fluid models and particle-dynamics models. In the two-fluid models, particles are treated In this section we briefly explain our model and how to simulate the dynamics of granular particles in fluid flows 15 or in Fig. 4 herein, we have seen that the kinetic energy is generated by bubbles with convec

Ichiki, Kengo

287

QUANTITY AND CAPACITY EXPANSION DECISIONS FOR ETHANOL IN NEBRASKA AND A MEDIUM SIZED PLANT.  

E-Print Network (OSTI)

??Corn-based ethanol is the leader of sustainable sources of energy in the United States due to the abundance of corn and the popularity of ethanol-gasoline (more)

Khoshnoud, Mahsa

2012-01-01T23:59:59.000Z

288

Electron-beam processing of kilogram quantities of iridium for radioisotope thermoelectric generator applications  

DOE Green Energy (OSTI)

Iridium alloys are used as fuel-cladding materials in radioisotope thermoelectric generators (RTGs). Hardware produced at the Oak Ridge National Laboratory (ORNL) has been used in Voyagers I and 2, Galilee, and Ulysses spacecraft. An integral part of the production of iridium-sheet metal involves electron-beam (EB) processing. These processes include the degassing of powder-pressed compacts followed by multiple meltings in order to purify 500-g buttons of Ir-0.3% W alloy. Starting in 1972 and continuing into 1992, our laboratory EB processing was Performed (ca. 1970) in a 60-kW (20 kV at 3 A), two-gun system. In 1991, a new 150-kW EB gun facility was installed to complement the older unit. This paper describes how the newly installed system was qualified for production of RTG developmental work is discussed that will potentially improve the existing process by utilizing the capabilities of the new EB system.

Huxford, T.J.; Ohriner, E.K.

1992-01-01T23:59:59.000Z

289

Electron-beam processing of kilogram quantities of iridium for radioisotope thermoelectric generator applications  

DOE Green Energy (OSTI)

Iridium alloys are used as fuel-cladding materials in radioisotope thermoelectric generators (RTGs). Hardware produced at the Oak Ridge National Laboratory (ORNL) has been used in Voyagers I and 2, Galilee, and Ulysses spacecraft. An integral part of the production of iridium-sheet metal involves electron-beam (EB) processing. These processes include the degassing of powder-pressed compacts followed by multiple meltings in order to purify 500-g buttons of Ir-0.3% W alloy. Starting in 1972 and continuing into 1992, our laboratory EB processing was Performed (ca. 1970) in a 60-kW (20 kV at 3 A), two-gun system. In 1991, a new 150-kW EB gun facility was installed to complement the older unit. This paper describes how the newly installed system was qualified for production of RTG developmental work is discussed that will potentially improve the existing process by utilizing the capabilities of the new EB system.

Huxford, T.J.; Ohriner, E.K.

1992-12-31T23:59:59.000Z

290

Quantities of Arsenic Within the State of Florida Completed on June 30, 2003  

E-Print Network (OSTI)

producers for electricity market bid offers and wind farm maintenance tasks. We will develop the forecasting by simulating possible wind power volatility scenarios combined with load forecasting errors and otherDevelopment and Deployment of an Advanced Wind Forecasting Technique Project Overview December 2008

Florida, University of

291

Impacts of large quantities of wind energy on the electric power system  

E-Print Network (OSTI)

Wind energy has been surging on a global scale. Significant penetration of wind energy is expected to take place in the power system, bringing new challenges because of the variability and uncertainty of this renewable ...

Yao, Yuan, S.M. Massachusetts Institute of Technology

2011-01-01T23:59:59.000Z

292

Designing a simple, robust, precision robotic platform for medium quantity production  

E-Print Network (OSTI)

A niche which has yet to be saturated in the growing market of educational and research robotic platforms is the mechanically-simple, electronically-powerful research robot. Useful in fields such as algorithm and artificial ...

Lieberman, Janet Samantha

2007-01-01T23:59:59.000Z

293

Laboratory development of sludge washing and alkaline leaching processes: Test plan for FY 1994  

Science Conference Proceedings (OSTI)

The US Department of Energy plans to vitrify (as borosilicate glass) the large volumes of high-level radioactive wastes at the Hanford site. To reduce costs, pretreatment processes will be used to reduce the volume of borosilicate glass required for disposal. Several options are being considered for the pretreatment processes: (1) sludge washing with water or dilute hydroxide: designed to remove most of the Na from the sludge, thus significantly reducing the volume of waste to be vitrified; (2) sludge washing plus caustic leaching and/or metathesis (alkaline sludge leaching): designed to dissolve large quantities of certain nonradioactive elements, such as Al, Cr and P, thus reducing the volume of waste even more; (3) sludge washing, sludge dissolution, and separation of radionuclides from the dissolved sludge solutions (advanced processing): designed to remove all radionuclides for concentration into a minimum waste volume. This report describes a test plan for work that will be performed in FY 1994 under the Sludge Washing and Caustic Leaching Studies Task (WBS 0402) of the Tank Waste Remediation System (TWRS) Pretreatment Project. The objectives of the work described here are to determine the effects of sludge washing and alkaline leaching on sludge composition and the physical properties of the washed sludge and to evaluate alkaline leaching methods for their impact on the volume of borosilicate glass required to dispose of certain Hanford tank sludges.

Rapko, B.M.; Lumetta, G.J.

1994-07-01T23:59:59.000Z

294

Drift Natural Convection and Seepage at the Yucca Mountain Repository  

E-Print Network (OSTI)

from nuclear weapons decommissioning, byproducts from thewaste from weapons decommissioning, and vitrified wastecosts projected through decommissioning, which would occur

Halecky, Nicholaus Eugene

2010-01-01T23:59:59.000Z

295

Technitium Management at the Hanford Site  

SciTech Connect

The Hanford tank waste contains -26,000 Ci of technetium-99 (Tc-99), the majority of which is in the supernate fraction. Tc-99 is a long-lived radionuclide with a half-life of -212,000 years and, in its predominant pertechnetate (TcO{sub 4}) fonn, is highly soluble and very mobile in the vadose zone and ultimately the groundwater. Tc-99 is identified as the major dose contributor (in groundwater) by past Hanford site performance assessments and therefore considered a key radionuclide of concern at Hanford. The United States Department of Energy (DOE) River Protection Project's (RPP) long-term Tc-99 management strategy is to immobitize the Tc-99 in a waste fonn that will retain the Tc-99 for many thousands of years. To achieve this, the RPP flowsheet will immobilize the majority of the Tc-99 as a vitrified low-activity waste product that will be ultimately disposed on site in the Integrated Disposal Facility. The Tc-99 will be released gradually from the glass at very low rates such that the groundwater concentrations at any point in time would be substantially below regulatory limits.The liquid secondary waste will be immobilized in a low-temperature matrix (cast stone) and the solid secondary waste will be stabilized using grout. Although the Tc-99 that is immobilized in glass will meet the release rate for disposal in IDF, a proportion is driven into the secondary waste stream that will not be vitrified and therefore presents a disposal risk. If a portion of the Tc-99 were to be removed from the Hanford waste inventory and disposed off-site, (e.g., as HLW), it could lessen a major constraint on LAW waste form performance, i.e., the requirement to retain Tc-99 over thousands of years and have a positive impact on the IDF Performance Assessment. There are several technologies available at various stages of technical maturity that can be employed for Tc-99 removal. The choice of technology and the associated efficacy of the technology are dependent on the chemical fonn of the technetium in the waste, the removal location in the tlowsheet. and the ultimate disposition path chosen for the technetium product. This paper will discuss the current plans for the management of the technetium present in the Hanford tank waste. It will present the risks associated with processing technetium in the current treatment tlowsheet and present potential mitigation opportunities, the status of available technetium removal technologies, the chemical speciation of technetium in the tank waste, and the available disposition paths and waste fonns for technetium containing streams.

Robbins, Rebecca A.

2013-08-15T23:59:59.000Z

296

IMPACT OF NOBLE METALS AND MERCURY ON HYDROGEN GENERATION DURING HIGH LEVEL WASTE PRETREATMENT AT THE SAVANNAH RIVER SITE  

DOE Green Energy (OSTI)

The Defense Waste Processing Facility (DWPF) at the Savannah River Site vitrifies radioactive High Level Waste (HLW) for repository internment. The process consists of three major steps: waste pretreatment, vitrification, and canister decontamination/sealing. HLW consists of insoluble metal hydroxides (primarily iron, aluminum, calcium, magnesium, manganese, and uranium) and soluble sodium salts (carbonate, hydroxide, nitrite, nitrate, and sulfate). The pretreatment process in the Chemical Processing Cell (CPC) consists of two process tanks, the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mix Evaporator (SME) as well as a melter feed tank. During SRAT processing, nitric and formic acids are added to the sludge to lower pH, destroy nitrite and carbonate ions, and reduce mercury and manganese. During the SME cycle, glass formers are added, and the batch is concentrated to the final solids target prior to vitrification. During these processes, hydrogen can be produced by catalytic decomposition of excess formic acid. The waste contains silver, palladium, rhodium, ruthenium, and mercury, but silver and palladium have been shown to be insignificant factors in catalytic hydrogen generation during the DWPF process. A full factorial experimental design was developed to ensure that the existence of statistically significant two-way interactions could be determined without confounding of the main effects with the two-way interaction effects. Rh ranged from 0.0026-0.013% and Ru ranged from 0.010-0.050% in the dried sludge solids, while initial Hg ranged from 0.5-2.5 wt%, as shown in Table 1. The nominal matrix design consisted of twelve SRAT cycles. Testing included: a three factor (Rh, Ru, and Hg) study at two levels per factor (eight runs), three duplicate midpoint runs, and one additional replicate run to assess reproducibility away from the midpoint. Midpoint testing was used to identify potential quadratic effects from the three factors. A single sludge simulant was used for all tests and was spiked with the required amount of noble metals immediately prior to performing the test. Acid addition was kept effectively constant except to compensate for variations in the starting mercury concentration. SME cycles were also performed during six of the tests.

Stone, M; Tommy Edwards, T; David Koopman, D

2009-03-03T23:59:59.000Z

297

Pulse Jet Mixer Overblow Testing for Assessment of Loadings During Multiple Overblows  

SciTech Connect

The U.S. Department of Energy (DOE) Office of River Protections Waste Treatment Plant (WTP) is being designed and built to pretreat and then vitrify a large portion of the wastes in Hanfords 177 underground waste storage tanks. The WTP consists of three primary facilities: pretreatment, low-activity waste (LAW) vitrification, and high-level waste (HLW) vitrification. The pretreatment facility will receive waste feed from the Hanford tank farms and separate it into 1) a high-volume, low-activity liquid stream stripped of most solids and radionuclides and 2) a much smaller volume of HLW slurry containing most of the solids and most of the radioactivity. Many of the vessels in the pretreatment facility will contain pulse jet mixers (PJMs) that will provide some or all of the mixing in the vessels. This technology was selected for use in so-called black cell regions of the WTP, where maintenance capability will not be available for the operating life of the WTP. PJM technology was selected for use in these regions because it has no moving mechanical parts that require maintenance. The vessels with the most concentrated slurries will also be mixed with air spargers and/or steady jets in addition to the mixing provided by the PJMs. This report contains the results of single and multiple PJM overblow tests conducted in a large, ~13 ft-diameter 15-ft-tall tank located in the high bay of the Pacific Northwest National Laboratory (PNNL) 336 Building test facility. These single and multiple PJM overblow tests were conducted using water and a clay simulant to bound the lower and upper rheological properties of the waste streams anticipated to be processed in the WTP. Hydrodynamic pressures were measured at a number of locations in the test vessel using an array of nine pressure sensors and four hydrophones. These measurements were made under normal and limiting vessel operating conditions (i.e., maximum PJM fluid emptying velocity, maximum and minimum vessel contents for PJM operation, and maximum and minimum rheological properties). Test data collected from the PJM overblow tests were provided to Bechtel National, Inc. (BNI) for assessing hydrostatic, dynamic, and acoustic pressure loadings on in-tank structures during 1) single overblows; 2) multiple overlapping overblows of two to four PJMs; 3) simultaneous overblows of pairs of PJMs.

Pfund, David M.; Bontha, Jagannadha R.; Michener, Thomas E.; Nigl, Franz; Yokuda, Satoru T.; Leigh, Richard J.; Golovich, Elizabeth C.; Baumann, Aaron W.; Kurath, Dean E.; Hoza, Mark; Combs, William H.; Fort, James A.; Bredt, Ofelia P.

2008-03-03T23:59:59.000Z

298

Pulse Jet Mixer Overblow Testing for Assessment of Loadings During Multiple Overblows  

SciTech Connect

The U.S. Department of Energy (DOE) Office of River Protections Waste Treatment Plant (WTP) is being designed and built to pretreat and then vitrify a large portion of the wastes in Hanfords 177 underground waste storage tanks. The WTP consists of three primary facilities: pretreatment, low-activity waste (LAW) vitrification, and high-level waste (HLW) vitrification. The pretreatment facility will receive waste feed from the Hanford tank farms and separate it into 1) a high-volume, low-activity liquid stream stripped of most solids and radionuclides and 2) a much smaller volume of HLW slurry containing most of the solids and most of the radioactivity. Many of the vessels in the pretreatment facility will contain pulse jet mixers (PJMs) that will provide some or all of the mixing in the vessels. This technology was selected for use in so-called black cell regions of the WTP, where maintenance capability will not be available for the operating life of the WTP. PJM technology was selected for use in these regions because it has no moving mechanical parts that require maintenance. The vessels with the most concentrated slurries will also be mixed with air spargers and/or steady jets in addition to the mixing provided by the PJMs. This report contains the results of single and multiple PJM overblow tests conducted in a large, ~13 ft-diameter 15-ft-tall tank located in the high bay of the Pacific Northwest National Laboratory (PNNL) 336 Building test facility. These single and multiple PJM overblow tests were conducted using water and a clay simulant to bound the lower and upper rheological properties of the waste streams anticipated to be processed in the WTP. Hydrodynamic pressures were measured at a number of locations in the test vessel using an array of nine pressure sensors and four hydrophones. These measurements were made under normal and limiting vessel operating conditions (i.e., maximum PJM fluid emptying velocity, maximum and minimum vessel contents for PJM operation, and maximum and minimum rheological properties). Test data collected from the PJM overblow tests were provided to Bechtel National, Inc. (BNI) for assessing hydrostatic, dynamic, and acoustic pressure loadings on in-tank structures during 1) single overblows; 2) multiple overlapping overblows of two to four PJMs; 3) simultaneous overblows of pairs of PJMs.

Pfund, David M.; Bontha, Jagannadha R.; Michener, Thomas E.; Nigl, Franz; Yokuda, Satoru T.; Leigh, Richard J.; Golovich, Elizabeth C.; Baumann, Aaron W.; Kurath, Dean E.; Hoza, Mark; Combs, William H.; Fort, James A.; Bredt, Ofelia P.

2009-07-20T23:59:59.000Z

299

RADIOACTIVE DEMONSTRATIONS OF FLUIDIZED BED STEAM REFORMING WITH ACUTAL HANFORD LOW ACTIVITY WASTES VERIFYING FBSR AS A SUPPLEMENTARY TREATMENT  

SciTech Connect

The U.S. Department of Energy's Office of River Protection is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the cleanup mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA). Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. Fluidized Bed Steam Reforming (FBSR) is one of the supplementary treatments being considered. FBSR offers a moderate temperature (700-750 C) continuous method by which LAW and other secondary wastes can be processed irrespective of whether they contain organics, nitrates/nitrites, sulfates/sulfides, chlorides, fluorides, and/or radio-nuclides like I-129 and Tc-99. Radioactive testing of Savannah River LAW (Tank 50) shimmed to resemble Hanford LAW and actual Hanford LAW (SX-105 and AN-103) have produced a ceramic (mineral) waste form which is the same as the non-radioactive waste simulants tested at the engineering scale. The radioactive testing demonstrated that the FBSR process can retain the volatile radioactive components that cannot be contained at vitrification temperatures. The radioactive and nonradioactive mineral waste forms that were produced by co-processing waste with kaolin clay in an FBSR process are shown to be as durable as LAW glass.

Jantzen, C.; Crawford, C.; Burket, P.; Bannochie, C.; Daniel, G.; Nash, C.; Cozzi, A.; Herman, C.

2012-01-12T23:59:59.000Z

300

Secondary Waste Form Down-Selection Data PackageFluidized Bed Steam Reforming Waste Form  

SciTech Connect

The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

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

2011-09-12T23:59:59.000Z

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

Adequacy of a Small Quantity Site RH-TRU Waste Program in Meeting Proposed WIPP Characterization Objectives  

SciTech Connect

The first remote-handled transuranic (RH-TRU) waste is expected to be permanently disposed of at the Waste Isolation Pilot Plant (WIPP) during Fiscal Year (FY) 2003. The first RH-TRU waste shipments are scheduled from the Battelle Columbus Laboratories (BCL) to WIPP in order to facilitate compliance with BCL Decommissioning Project (BCLDP) milestones. Milestones requiring RH-TRU waste containerization and removal from the site by 2004 in order to meet a 2006 site closure goal, established by Congress in the Defense Facilities Closure Projects account, necessitated the establishment and implementation of a site-specific program to direct the packaging of BCLDP RH-TRU waste prior to the finalization of WIPP RH-TRU waste characterization requirements. The program was designed to collect waste data, including audio and videotape records of waste packaging, such that upon completion of waste packaging, comprehensive data records exist from which compliance with final WIPP RH-TRU waste characterization requirements can be demonstrated. With the BCLDP data records generated to date and the development by the U.S. Department of Energy (DOE)-Carlsbad Field Office (CBFO) of preliminary documents proposing the WIPP RH-TRU waste characterization program, it is possible to evaluate the adequacy of the BCLDP program with respect to meeting proposed characterization objectives. The BCLDP characterization program uses primarily acceptable knowledge (AK) and visual examination (VE) during waste packaging to characterize RH-TRU waste. These methods are used to estimate physical waste parameters, including weight percentages of metals, cellulosics, plastics, and rubber in the waste, and to determine the absence of prohibited items, including free liquids. AK combined with computer modeling is used to estimate radiological waste parameters, including total activity on a waste container basis, for the majority of BCLDP RH-TRU waste. AK combined with direct analysis is used to characterize radiological parameters for the small populations of the RH-TRU waste generated by the BCLDP. All characterization based on AK is verified. Per its design for comprehensive waste data collection, the BCLDP characterization program using AK and waste packaging procedures, including VE during packaging, meets the proposed WIPP RH-TRU waste characterization objectives. The conservative program design implemented generates certification data that will be adequate to meet any additional program requirements that may be imposed by the CBFO.

Biedscheid, J.; Stahl, S.; Devarakonda, M.; Peters, K.; Eide, J.

2002-02-26T23:59:59.000Z

302

Measurements and Characterization of Neutron and Gamma Dose Quantities in the Vicinity of an Independent Spent Fuel Storage Installation  

SciTech Connect

As part of the decommissioning of the Maine Yankee Atomic Power Company (MYAPCo) nuclear power plant, the spent nuclear fuel is being temporarily stored in a dry cask storage facility on a portion of the original licensed property. Each of the spent nuclear fuel (SNF) storage casks hold approximately 25 spent fuel assemblies. Additional storage casks for the greater-than-Class C waste (GTCC) are also used. This waste is contained in 64 casks (60 SNF, 4 GTCC), each of which contain a substantial amount of concrete for shielding and structural purposes. The vertical concrete casks (VCCs) are typically separated by a distance of 4 and 6 feet. The storage casks are effective personnel radiation shields for most of the gamma and neutron radiation emitted from the fuel. However measurable gamma and neutron radiation levels are present in the vicinity of the casks. In order to establish a controlled area boundary around the facility such that a member of the public annual dose level of 0.25-mSv could be demonstrated, measurements of gamma and neutron dose equivalents were conducted. External gamma exposure rates were measured with a Pressurized Ion Chamber (PIC). Neutron absorbed dose and dose equivalent rates were measured with a Rossi-type tissue equivalent proportional counter (TEPC). Both gamma and neutron measurements were made at increasing distances from the facility as well as at a background location. The results of the measurements show that the distance to the 0.25-mSv per year boundary for 100% occupancy conditions varies from 321 feet to 441 feet from the geometric center of the storage pads, depending on the direction from the pad. For the TEPC neutron measurements, the average quality factor from the facilities was approximately 7.4. This quality factor compares well with the average quality factor of 7.6 that was measured during a calibration performed with a bare Cf-252 source. (authors)

Darois, E.L.; Keefer, D.G.; Plazeski, P.E. [Radiation Safety and Control Services, 91 Portsmouth Avenue, Stratham, NH 03885 (United States); Connell, J. [Maine Yankee Atomic Power Company, Wiscasset, ME 04568 (United States)

2006-07-01T23:59:59.000Z

303

The influence of high quantity of fly ash on reducing the expansion due to ASR in the presence of alkalis  

E-Print Network (OSTI)

A testing program was devised to study the role of high volume fly ash (HVFA) in reducing the expansion caused by alkali-silica reaction (ASR). A series of modified ASTM C 1260 tests were performed, where the replacement of cement by Class F fly ash was 58% by mass of cement. A reactive siliceous aggregate was used. The influence of inherent alkalis in cement to the reaction was also studied. The test results confirm that HVFA significantly helps in controlling expansion caused by ASR. The test period was extended to 28 days to assess if more reproducible results can be obtained. The results indicate that reducing the alkalinity of the sodium hydroxide solution by 50%, to 0.5N is sufficient to determine the potential reactivity of aggregates. The reduction of alkalinity of sodium hydroxide to 0.25N, however, produced results, which were beyond interpretation. Concrete using High Volume Fly Ash was tested for strength to ascertain if the reactive aggregates or the percentage of internal alkalis in the cement influenced the strength. This report discusses the test results for only part of a broader research program in progress at the Texas Transportation Institute, Texas A&M University.

Mohidekar, Saleel D.

2000-01-01T23:59:59.000Z

304

Estimation of rotor angles of synchronous machines using artificial neural networks and local PMU-based quantities  

Science Conference Proceedings (OSTI)

This paper investigates a possibility for estimating rotor angles in the time frame of transient (angle) stability of electric power systems, for use in real-time. The proposed dynamic state estimation technique is based on the use of voltage and current ... Keywords: Electric power systems, Multilayer perceptrons, Phasor measurement units, Transient stability monitoring and control

Alberto Del Angel; Pierre Geurts; Damien Ernst; Mevludin Glavic; Louis Wehenkel

2007-10-01T23:59:59.000Z

305

2. Annual Quantities of Land Clearing Debris Generated and Used (1) Sectors that generate Land Clearing Debris:  

E-Print Network (OSTI)

Land clearing debris is defined as growing stock and other timber sources cut or otherwise destroyed in the process of converting forest land to non-forest uses. 1 Growing stock that is removed in silvicultural operations such as pre-commercial thinning is also included in this definition. Land clearing debris is typically in the form of tree tops and branches, trees cut or knocked down and left on site, and stumps. In non-forested areas, such as grasslands and desert, land clearing debris may include soil, rocks, and shrubs, although fuel is primarily derived from previously forested areas.

unknown authors

2010-01-01T23:59:59.000Z

306

West Texas high school agriscience teachers' knowledge, confidence, and attitudes towards teaching water quantity-related topics.  

E-Print Network (OSTI)

??As the nations population grows, the water supply is depleting. Since agricultural education plays a large role in many Texas high schools, it is important (more)

Miller, Pamela Marie

2006-01-01T23:59:59.000Z

307

Investigation of vesicular arbuscular mycorrhizal (VAM) on yield quantity and quality of sorghum cultivars under irrigation in arid area  

E-Print Network (OSTI)

the control. Meanwhile, sorghum cultivars showed differentfor high yielding in sorghum when sowing in low irrigationcondition. We exposed sorghum to various treatments just

Moussavinik, Mohsen; Mehraban, Ahmad

2009-01-01T23:59:59.000Z

308

DOE O 462.1 Admin Chg 1, Import and Export of Category 1 and 2 Radioactive Sources Aggregated Quantities  

Directives, Delegations, and Requirements

To formalize relevant guidance contained in the International Atomic Energy Agency (IAEA) CODEOC 2004, Code of Conduct on the Safety and Security of ...

2008-11-10T23:59:59.000Z

309

ESTIMATING HIGH LEVEL WASTE MIXING PERFORMANCE IN HANFORD DOUBLE SHELL TANKS  

SciTech Connect

The ability to effectively mix, sample, certify, and deliver consistent batches of high level waste (HLW) feed from the Hanford double shell tanks (DSTs) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. The Department of Energy's (DOE's) Tank Operations Contractor (TOC), Washington River Protection Solutions (WRPS) is currently demonstrating mixing, sampling, and batch transfer performance in two different sizes of small-scale DSTs. The results of these demonstrations will be used to estimate full-scale DST mixing performance and provide the key input to a programmatic decision on the need to build a dedicated feed certification facility. This paper discusses the results from initial mixing demonstration activities and presents data evaluation techniques that allow insight into the performance relationships of the two small tanks. The next steps, sampling and batch transfers, of the small scale demonstration activities are introduced. A discussion of the integration of results from the mixing, sampling, and batch transfer tests to allow estimating full-scale DST performance is presented.

THIEN MG; GREER DA; TOWNSON P

2011-01-13T23:59:59.000Z

310

ENHANCED CHEMICAL CLEANING: A NEW PROCESS FOR CHEMICALLY CLEANING SAVANNAH RIVER WASTE TANKS  

SciTech Connect

The Savannah River Site (SRS) has 49 high level waste (HLW) tanks that must be emptied, cleaned, and closed as required by the Federal Facilities Agreement. The current method of chemical cleaning uses several hundred thousand gallons per tank of 8 weight percent (wt%) oxalic acid to partially dissolve and suspend residual waste and corrosion products such that the waste can be pumped out of the tank. This adds a significant quantity of sodium oxalate to the tanks and, if multiple tanks are cleaned, renders the waste incompatible with the downstream processing. Tank space is also insufficient to store this stream given the large number of tanks to be cleaned. Therefore, a search for a new cleaning process was initiated utilizing the TRIZ literature search approach, and Chemical Oxidation Reduction Decontamination--Ultraviolet (CORD-UV), a mature technology currently used for decontamination and cleaning of commercial nuclear reactor primary cooling water loops, was identified. CORD-UV utilizes oxalic acid for sludge dissolution, but then decomposes the oxalic acid to carbon dioxide and water by UV treatment outside the system being treated. This allows reprecipitation and subsequent deposition of the sludge into a selected container without adding significant volume to that container, and without adding any new chemicals that would impact downstream treatment processes. Bench top and demonstration loop measurements on SRS tank sludge stimulant demonstrated the feasibility of applying CORD-UV for enhanced chemical cleaning of SRS HLW tanks.

Ketusky, E; Neil Davis, N; Renee Spires, R

2008-01-17T23:59:59.000Z

311

Seagate Crystal Reports - sum6.  

Office of Environmental Management (EM)

Shipping and Shipping and Receiving Activity (Sum-6) Current Year: 2000 Receiving Site: Hanford Shipping Site HLW HL W -V i trified TRU M L L W LLW OTHER* SNF** Quantity (m 3) Quantity (m 3) Quantity ( m 3) Quantity (m 3) Quantity (m 3) Quantity ( m 3) Quantity (M THM ) Ames Lab 0.000 0.000 0.000 0.000 5.460 0.000 0.0000 Argonne-E 0.000 0.000 0.000 0.000 1,049.800 0.000 0.0000 Bettis 0.000 0.000 0.000 0.000 11.680 0.000 0.0000 Brookhaven 0.000 0.000 0.000 0.000 55.070 0.000 0.0000 Columbus 0.000 0.000 0.000 0.000 156.070 0.000 0.0000 EnergyTech 0.000 0.000 0.000 0.000 41.780 0.000 0.0000 Fermi 0.000 0.000 0.000 0.000 42.840 0.000 0.0000 GenAtomics 0.000 0.000 0.000 0.000 164.030 0.000 0.0000 Lawr-Berk 0.000 0.000 0.000 0.000 12.220 0.000 0.0000 NavRctrFac 0.000 0.000 0.000 16.000 0.000 0.000 0.0000

312

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

SciTech Connect

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

Ketusky, Edward; Spires, Renee; Davis, Neil

2009-02-11T23:59:59.000Z

313

Risk-informing decisions about high-level nuclear waste repositories  

E-Print Network (OSTI)

Performance assessments (PAs) are important sources of information for societal decisions in high-level radioactive waste (HLW) management, particularly in evaluating safety cases for proposed HLW repository development. ...

Ghosh, Suchandra Tina, 1973-

2004-01-01T23:59:59.000Z

314

HANFORD RIVER PROTECTION PROJECT ENHANCED MISSION PLANNING THROUGH INNOVATIVE TOOLS LIFECYCLE COST MODELING AND AQUEOUS THERMODYNAMIC MODELING - 12134  

SciTech Connect

Two notable modeling efforts within the Hanford Tank Waste Operations Simulator (HTWOS) are currently underway to (1) increase the robustness of the underlying chemistry approximations through the development and implementation of an aqueous thermodynamic model, and (2) add enhanced planning capabilities to the HTWOS model through development and incorporation of the lifecycle cost model (LCM). Since even seemingly small changes in apparent waste composition or treatment parameters can result in large changes in quantities of high-level waste (HLW) and low-activity waste (LAW) glass, mission duration or lifecycle cost, a solubility model that more accurately depicts the phases and concentrations of constituents in tank waste is required. The LCM enables evaluation of the interactions of proposed changes on lifecycle mission costs, which is critical for decision makers.

PIERSON KL; MEINERT FL

2012-01-26T23:59:59.000Z

315

Summary - Tank 48 at the Savannah River Site  

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

Tank 48 Tank 48 ETR Report Date: August 2006 ETR-2 United States Department of Energy Office of Environmental Management (DOE-EM) External Technical Review of Tank 48 at the Savannah River Site (SRS) Why DOE-EM Did This Review Tank 48 is a 1.3 million gallon tank with full secondary containment, located and interconnected within the SRS tank system that will play a very important role in removal and processing of high-level waste (HLW) in the years ahead. However, the tank is currently isolated from the system and unavailable for use, because its contents. It contains approximately 250,000 gallons of salt solution containing Cesium-137 and other radioisotopes which are contaminated with significant quantities of tetraphenylborate (TPB), a material which

316

Savannah River Site - Tank 48 SRS Review Report  

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

ETR-2 United States Department of Energy Office of Environmental Management (DOE-EM) External Technical Review of Tank 48 at the Savannah River Site (SRS) Why DOE-EM Did This Review Tank 48 is a 1.3 million gallon tank with full secondary containment, located and interconnected within the SRS tank system that will play a very important role in removal and processing of high-level waste (HLW) in the years ahead. However, the tank is currently isolated from the system and unavailable for use, because its contents. It contains approximately 250,000 gallons of salt solution containing Cesium-137 and other radioisotopes which are contaminated with significant quantities of tetraphenylborate (TPB), a material which can release benzene vapor to the tank head space in

317

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

SciTech Connect

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

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

1996-05-01T23:59:59.000Z

318

MacroscoMacroscopic Cracking Determination in LaBS Glasspic Cracking Determination in LaBS Glass  

SciTech Connect

The DOE/EM plans to conduct the Plutonium Vitrification Project at the Savannah River Site (SRS). An important part of this project is to reduce the attractiveness of the plutonium by fabricating a plutonium glass form and immobilizing the Pu form within the high level waste (HLW) glass prepared in the Defense Waste Processing Facility (DWPF). This requires that a project schedule that is consistent with EM plans for DWPF and cleanup of the SRS be developed. Critical inputs to key decisions in the vitrification project schedule are near-term data that will increase confidence that the lanthanide borosilicate (LaBS) glass product is suitable for disposal in the Yucca Mountain Repository. A workshop was held on April 28, 2005 at Bechtel SAIC Company (BSC) facility in Las Vegas, NV to define the near term data needs. Dissolution rate data and the fate of plutonium oxide and the neutron absorbers during the dissolution process were defined as key data needs. A suite of short-term tests were defined at the workshop to obtain the needed data. The objectives of these short-term tests are to obtain data that can be used to show that the dissolution rate of a LaBS glass is acceptable and to show that the extent of Pu separation from neutron absorbers, as the glass degrades and dissolves, is not likely to lead to criticality concerns. An additional data need was identified regarding the degree of macroscopic cracking and/or voiding that occurs during processing of the Pu glass waste form and subsequent pouring of HLW glass in the DWPF. A final need to evaluate new frit formulations that may increase the durability of the plutonium glass and/or decrease the degree to which neutron absorbers separate from the plutonium during dissolution was identified. This task plan covers the need to evaluate the degree of macroscopic cracking and/or voiding that occurs during processing of the Vitrified Plutonium Waste Form (i.e. the can-in-canister configuration containing the vitrified Pu product). Separate task plans were developed for Pu glass performance testing of the current baseline LaBS glass composition and development of alternative frit formulations. Recent results from Pressurized Unsaturated Flow (PUF) testing showed the potential separation of Pu from Gd during the glass dissolution process [3]. Post-test analysis of the LaBS glass from a 6-year PUF test showed a region where Pu had apparently accumulated in a Pu-bearing disk-like phase that had become separated from neutron absorber (Gd). It should be noted that this testing was conducted on the early LaBS Frit A glass composition that was devoid of HfO{sub 2} as a neutron absorber. PUF testing is currently being initiated using the LaBS Frit B composition that contains HfO{sub 2}. The potential for fissile material and neutron absorber separation is a criticality risk for the repository. The surface area that is available for leaching (i.e. due to the degree of cracking or voiding within the Pu glass cylinder) is a factor in modeling the amount of fissile material and neutron absorber released during the dissolution process. A mathematical expression for surface area is used in the Total Systems Performance Assessment (TSPA) performed by BSC personnel. Specifically, the surface area available for leaching is being used in current external criticality assessments. The planned processing steps for producing a VPWF assembly involves processing Pu feed and LaBS frit to produce a can of Pu LaBS glass, packaging this can into a second can (i.e. bagless transfer) for removal from the glovebox processing environment, placing a series of bagless transfer cans into a DWPF canister, and pouring HLW glass into the DWPF canister to encapsulate bagless transfer cans. The objective of this task is to quantify the degree of cracking and/or voiding that will occur during the processing of the VPWF.

Marra, James

2005-08-01T23:59:59.000Z

319

Amorphous Aluminum AlloysSynthesis and Stability  

Science Conference Proceedings (OSTI)

Recent innovations in metallic glasses have led to new alloy classes that may be vitrified and a re-examination of the key alloying factors influencing glass...

320

DRAFT Forensics@NIST 2012 Symposium Abstracts by Day ...  

Science Conference Proceedings (OSTI)

... vitrified composite (UVC) doped with enriched uranium to test ... For example, the highly volatile compound TATP ... provides an initially high amount of ...

2012-09-12T23:59:59.000Z

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

A Characteristics-Based Approach to Radioactive Waste Classification in Advanced Nuclear Fuel Cycles  

E-Print Network (OSTI)

LLW-?GTCC GTCC in NRC criteria. Vitrified glass; Matheson, Jim. Letter to NRC Chairman Dale E California Press, NRC 2008 Nuclear Regulatory

Djokic, Denia

2013-01-01T23:59:59.000Z

322

Vitrification and Glass Characterization for Nuclear Materials Disposal  

Science Conference Proceedings (OSTI)

Oct 20, 2011 ... One significant limitation to waste loading in glass for Hanford .... to the high level sludge vitrified at the Defense Waste Processing Facility.

323

Materials for Nuclear Power: Digital Resource Center - ABSTRACT ...  

Science Conference Proceedings (OSTI)

May 22, 2007... whose overall composition is representative of the high chrome oxide wastes at Hanford WA USA, was easily vitrified in a phosphate glass at...

324

Defense High Level Waste Disposal Container System Description Document  

Science Conference Proceedings (OSTI)

The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents.

N. E. Pettit

2001-07-13T23:59:59.000Z

325

Idaho National Engineering Laboratory High-Level Waste Roadmap. Revision 2  

SciTech Connect

The Idaho National Engineering Laboratory (INEL) High-Level Waste (HLW) Roadmap takes a strategic look at the entire HLW life-cycle starting with generation, through interim storage, treatment and processing, transportation, and on to final disposal. The roadmap is an issue-based planning approach that compares ``where we are now`` to ``where we want and need to be.`` The INEL has been effectively managing HLW for the last 30 years. Calcining operations are continuing to turn liquid HLW into a more manageable form. Although this document recognizes problems concerning HLW at the INEL, there is no imminent risk to the public or environment. By analyzing the INEL current business operations, pertinent laws and regulations, and committed milestones, the INEL HLW Roadmap has identified eight key issues existing at the INEL that must be resolved in order to reach long-term objectives. These issues are as follows: A. The US Department of Energy (DOE) needs a consistent policy for HLW generation, handling, treatment, storage, and disposal. B. The capability for final disposal of HLW does not exist. C. Adequate processes have not been developed or implemented for immobilization and disposal of INEL HLW. D. HLW storage at the INEL is not adequate in terms of capacity and regulatory requirements. E. Waste streams are generated with limited consideration for waste minimization. F. HLW is not adequately characterized for disposal nor, in some cases, for storage. G. Research and development of all process options for INEL HLW treatment and disposal are not being adequately pursued due to resource limitations. H. HLW transportation methods are not selected or implemented. A root-cause analysis uncovered the underlying causes of each of these issues.

1993-08-01T23:59:59.000Z

326

SLUDGE BATCH 4 SIMULANT FLOWSHEET STUDIES: PHASE II RESULTS  

DOE Green Energy (OSTI)

The Defense Waste Processing Facility (DWPF) will transition from Sludge Batch 3 (SB3) processing to Sludge Batch 4 (SB4) processing in early fiscal year 2007. Tests were conducted using non-radioactive simulants of the expected SB4 composition to determine the impact of varying the acid stoichiometry during the Sludge Receipt and Adjustment Tank (SRAT) process. The work was conducted to meet the Technical Task Request (TTR) HLW/DWPF/TTR-2004-0031 and followed the guidelines of a Task Technical and Quality Assurance Plan (TT&QAP). The flowsheet studies are performed to evaluate the potential chemical processing issues, hydrogen generation rates, and process slurry rheological properties as a function of acid stoichiometry. Initial SB4 flowsheet studies were conducted to guide decisions during the sludge batch preparation process. These studies were conducted with the estimated SB4 composition at the time of the study. The composition has changed slightly since these studies were completed due to changes in the sludges blended to prepare SB4 and the estimated SB3 heel mass. The following TTR requirements were addressed in this testing: (1) Hydrogen and nitrous oxide generation rates as a function of acid stoichiometry; (2) Acid quantities and processing times required for mercury removal; (3) Acid quantities and processing times required for nitrite destruction; and (4) Impact of SB4 composition (in particular, oxalate, manganese, nickel, mercury, and aluminum) on DWPF processing (i.e. acid addition strategy, foaming, hydrogen generation, REDOX control, rheology, etc.).

Stone, M; David Best, D

2006-09-12T23:59:59.000Z

327

PJM Controller Testing with Prototypic PJM Nozzle Configuration  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy (DOE) Office of River Protections Waste Treatment Plant (WTP) is being designed and built to pre-treat and then vitrify a large portion of the wastes in Hanfords 177 underground waste storage tanks. The WTP consists of three primary facilitiespretreatment, low-activity waste (LAW) vitrification, and high-level waste (HLW) vitrification. The pretreatment facility will receive waste piped from the Hanford tank farms and separate it into a high-volume, low-activity liquid stream stripped of most solids and radionuclides and a much smaller volume of HLW slurry containing most of the solids and most of the radioactivity. Many of the vessels in the pretreatment facility will contain pulse jet mixers (PJM) that will provide some or all of the mixing in the vessels. Pulse jet mixer technology was selected for use in black cell regions of the WTP, where maintenance cannot be performed once hot testing and operations commence. The PJMs have no moving mechanical parts that require maintenance. The vessels with the most concentrated slurries will also be mixed with air spargers and/or steady jets in addition to the mixing provided by the PJMs. Pulse jet mixers are susceptible to overblows that can generate large hydrodynamic forces, forces that can damage mixing vessels or their internal parts. The probability of an overblow increases if a PJM does not fill completely. The purpose of the testing performed for this report was to determine how reliable and repeatable the primary and safety (or backup) PJM control systems are at detecting drive overblows (DOB) and charge vessel full (CVF) conditions. Testing was performed on the ABB 800xA and Triconex control systems. The controllers operated an array of four PJMs installed in an approximately 13 ft diameter 15 ft tall tank located in the high bay of the Pacific Northwest National Laboratory (PNNL) 336 Building test facility. The PJMs were fitted with 4 inch diameter discharge nozzles representative of the nozzles to be used in the WTP. This work supplemented earlier controller tests done on PJMs with 2 inch nozzles (Bontha et al. 2007). Those earlier tests enabled the selection of appropriate pressure transmitters with associated piping and resulted in an alternate overblow detection algorithm that uses data from pressure transmitters mounted in a water flush line on the PJM airlines. Much of that earlier work was only qualitative, however, due to a data logger equipment failure that occurred during the 2007 testing. The objectives of the current work focused on providing quantitative determinations of the ability of the BNI controllers to detect DOB and CVF conditions. On both control systems, a DOB or CVF is indicated when the values of particular internal functions, called confidence values, cross predetermined thresholds. There are two types of confidence values; one based on a transformation of jet pump pair (JPP) drive and suction pressures, the other based on the pressure in the flush line. In the present testing, we collected confidence levels output from the ABB and Triconex controllers. These data were analyzed in terms of the true and noise confidence peaks generated during multiple cycles of DOB and CVF events. The distributions of peak and noise amplitudes were compared to see if thresholds could be set that would enable the detection of DOB and CVF events at high probabilities, while keeping false detections to low probabilities. Supporting data were also collected on PJM operation, including data on PJM pressures and levels, to provide direct experimental evidence of when PJMs were filling, full, driving, or overblowing.

Bontha, Jagannadha R.; Nigl, Franz; Weier, Dennis R.; Leigh, Richard J.; Johnson, Eric D.; Wilcox, Wayne A.; Pfund, David M.; Baumann, Aaron W.; Wang, Yeefoo

2009-08-21T23:59:59.000Z

328

S. J. Cotter, F. R. O'Donnell, and P. A. Scofield Activities on the ORR have the potential to release small quantities of radionuclides and hazardous  

E-Print Network (OSTI)

#12;#12;Contents ORNL/PPA-2006/1 Oak Rdge Natonal Laboratory LABORATORY DIRECTED RESEARCH Contract DE-AC05-00OR22725 #12; FY 2004 ORNL Laboratory Drected Research and Development Annual Report #12.................................................................................... 88 #12;v FY 2004 ORNL Laboratory Drected Research and Development Annual Report Development

Pennycook, Steve

329

DETERMINATION OF THE QUANTITY OF I-135 RELEASED FROM THE AGR-1 TEST FUELS AT THE END OF ATR OPERATING CYCLE 138B  

SciTech Connect

The AGR-1 experiment is a multiple fueled-capsule irradiation experiment being conducted in the Advanced Test Reactor (ATR) in support of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and ended with shutdown of the reactor for a brief outage on February 10, 2007 at 0900. The AGR-1 experiment will continue cyclical irradiation for about 2.5 years. In order to allow estimation of the amount of radioiodine released during the first cycle, purge gas flow to all capsules continued for about 4 days after reactor shutdown. The FPMS data acquired during part of that shutdown flow period has been analyzed to elucidate the level of 135I released during the operating cycle.

J. K. Hartwell; D. M. Scates; J. B. Walter; M. W. Drigert

2007-05-01T23:59:59.000Z

330

Short Communication: A simple and better algorithm to solve the vendor managed inventory control system of multi-product multi-constraint economic order quantity model  

Science Conference Proceedings (OSTI)

This research presents an alternative heuristic algorithm to solve the vendor management inventory system with multi-product and multi-constraint based on EOQ with backorders considering two classical backorders costs: linear and fixed. For this type ... Keywords: EOQ, Genetic algorithms, Heuristic algorithms, Nonlinear integer programming, Vendor management inventory

Leopoldo Eduardo Crdenas-Barrn; Gerardo Trevio-Garza; Hui Ming Wee

2012-02-01T23:59:59.000Z

331

Composition and quantities of retained gas measured in Hanford waste tanks 241-U-103, S-106, BY-101, and BY-109  

DOE Green Energy (OSTI)

This report provides the results obtained for the single-shell tanks (SSTs) sampled with the Retained Gas Sampler (RGS) during 1997: Tanks 241-U-103, 241-S-106, 241-BY-101, and 241-BY-109. The RGS is a modified version of the core sampler used at Hanford. It is designed specifically to be used in concert with the gas extraction equipment in the hot cell to capture and extrude a gas-containing waste sample in a hermetically sealed system. The four tanks represent several different types of flammable gas SSTs. Tank U-103 is on the Flammable Gas Watch List (FGWL) and is one of the highest-priority group of SSTs that show evidence of significant gas retention. Tank S-106, though not a FGWL tank, has a uniquely high barometric pressure response and continuing rapid surface level rise, indicating a large and increasing volume of retained gas. Tanks BY-101 and BY-109 are not on the FGWL but were chosen to test the effect of recent salt-well pumping on gas retention. Section 2 of this report provides an overview of the process by which retained gases in the Hanford tanks are sampled and analyzed. A detailed description of the procedure used to reduce and analyze the data is provided in Section 3. Tank-by-tank results are covered in Section 4 (with the data presented in the order in which the tanks were sampled), and an RGS system performance overview is given in Section 5. Section 6 presents conclusions from these analyses and recommendations for further research. The cited references are listed in Section 7. Appendix A describes the procedures used to extract gas and ammonia from the samples, Appendix B contains detailed laboratory data from each of the tanks, and Appendix C gives field sampling data.

Mahoney, L.A.; Antoniak, Z.I.; Bates, J.M.

1997-12-01T23:59:59.000Z

332

Effect of Long-term Lime and Potassium Applications on Quantity-Intensity (Q/I) Relationships in Sandy Soil1  

E-Print Network (OSTI)

in Sandy Soil1 D. L. SPARKS AND W. C. LiEBHARDT2 ABSTRACT The effects of long-term lime and K applications on quan- tity-intensity (Q/I) relationships were investigated on the Ap and B21t horizons of a Kalmia soil, and chloritized ver- miculite. Soil pH and exchangeable bases increased with depth and with lime additions

Sparks, Donald L.

333

Understanding the Chemistry of the Actinides in HL Waste Tank Systems: Actinide Speciation in Oxalic Acid Solutions in the Presence of Significant Quantities of Aluminum, Iron, and Manganese  

SciTech Connect

The overall goal of this research plan is to provide a thermodynamic basis for describing actinide speciation over a range of tank-like conditions, including elevated temperature, elevated OH- concentrations, and the presence of various organic ligands. With support from DOE?s EMSP program, we have made significant progress towards measuring thermodynamic parameters for actinide complexation as a function of temperature. We have used the needs of the ESP modelers to guide our work to date, and we have made important progress defining the effect of temperature for actinide complexation by organic, and for hydrolysis of the hexa- and pentvalent oxidation states.

Clark, Sue

2006-07-30T23:59:59.000Z

334

Activities on ORR have the potential to release small quantities of radionuclides and hazardous chemicals to the environment. These releases could expose members of the public  

E-Print Network (OSTI)

to radiation from nuclides deposited inside the body are called internal exposures. This distinction these potential exposure pathways combined was estimated to be about 3 mrem. DOE O 458.1, Radiation Protection Exposures to radiation from nuclides located outside the body are called external exposures; exposures

Pennycook, Steve

335

Comparing Paleoclassical-Based Pedestal Model Predictions of Electron Quantities to Measured DIII-D H-mode Profiles (A27160)  

E-Print Network (OSTI)

Proc. Of 13th International Workshop On H-mode Physics And Transport, Oxford, United Kingdom, 2011; To Be Published In Nucl. Fusion13th International Workshop on H-mode Physics and Transport Barriers Oxford, UK, 2011999619117

Smith, S.P.

2011-12-15T23:59:59.000Z

336

2012 Annual Workforce Analysis and Staffing Plan Report - West Valley Demonstration Project  

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

ATTACHMENT ATTACHMENT 1 Annual Workforce Analysis and Staffing Plan Report As ofDecember 31, 2012 Reporting Office: West Valley Demonstration Project Section 1: Current Mission(s) of the Organization and Potential Changes The mission of the WVDP as defined by the West Valley Demonstration Project Act (Public Law 96-368) is to accomplish five activities: 1) solidify high-level radioactive waste (HLW), 2) develop containers suitable for permanent disposal of the HLW, 3) transport the HLW to a Federal repository for permanent disposal, 4) dispose of low-level and transuranic waste produced by the solidification of the HLW, and 5) decontaminate and decommission the HLW tanks and facilities, materials and hardware used to solidify the HLW. DOE expects to accomplish these WVDP activities through proactive leadership, management, and implementation of safe and environmentally sound operations.

337

Microsoft PowerPoint - 10-04 Sundar Technology Needs for WTP Simulants - PSSundar.ppt  

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

Needs for WTP Simulants Needs for WTP Simulants P. S. Sundar Process Technology - Plant Operations Div Waste Treatment Plant Project November 17, 2010 Bechtel National, Inc. Print Close Technology Needs for WTP Simulants 2 Agenda * Major simulant requirements of WTP Project and the associated challenges Bechtel National, Inc. Close Print Technology Needs for WTP Simulants 3 Simplified Process Flowsheet IHLW ILAW LAW Feed HLW Feed HLW Recycles LAW Recycles Bechtel National, Inc. Close Print Technology Needs for WTP Simulants 4 Simulant Needs * Commissioning Simulants - As received and pretreated LAW supernatants - As received HLW sludge - Pretreated HLW sludge - Vitrification recycle streams

338

Kinetic Studies of Nepheline Formation in Advanced Silicate ...  

Nepheline Phase Field Recent developments in advanced silicate glasses suggest that step function improvements in waste loadings of HLW ... Historic T-T-T diagrams ...

339

Hazards control progress report No. 51, July--December 1975  

SciTech Connect

Progress is reported on research projects in the fields of radiation protection, industrial hygiene, instrument development, fire safety, decontamination, and environmental protection. (HLW)

Crites, T.R. (comp.)

1976-02-16T23:59:59.000Z

340

DOE/ID-Number  

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

This includes the storage, transportation, and disposal of low level waste (LLW), used nuclear fuel (UNF), and high level waste (HLW). The Office of Fuel Cycle Technologies...

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

Microsoft PowerPoint - 6-04 Bowan-Smith_r1.ppt  

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

system Completed HLW Melters Completed LAW Melter Print Close AJHM Program * Goal: Increase glass production rate for existing melter foot-print with minimal impact on WTP:...

342

EAC Plan  

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

Facility Mixes waste feed with glass formers Two HLW melters - 3 MTd per melter Remote handling equipment 480 canisters average per year Canister handling Offgas...

343

System Plan Revision 4  

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

@ 90, Hanford 2 pumps @ 180 * Different Feed Receipt Constraints - SRS "good glass", Hanford RCRA and rheology DST Mixing and HLW Certification 19 52109 Programmatic...

344

U.S. Department of Energy Categorical Exclusion ...  

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

Hanford WTP HLW Canister Decontamination Process Stream Simulant Development Savannah River Site AikenAikenSouth Carolina A simulant will be developed of the canister...

345

app_d  

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

53 DOEEIS-0287 Idaho HLW & FD EIS - New Information - Document 58, COGEMA, Inc. (Rhonnie Smith), Idaho Falls, ID Page 1 of 13 Document 57, Studsvik, Inc. (Thomas Oliver),...

346

Environmental Sciences Division annual progress report for period ending September 30, 1978. Environmental Sciences Division publication No. 1280  

DOE Green Energy (OSTI)

Separate abstracts were prepared for the ten sections of the report. The report also includes lists of publications, theses, professional activities, and organization charts. (HLW)

Auerbach, S.I.; Reichle, D.E.; Struxness, E.G.

1979-04-01T23:59:59.000Z

347

High Level Waste System at SRS  

Tank Under Construction Tanks are built at grade and then backfilled with dirt to provide ... Hanford discussion. 2005-01-19 2005-01-19 HLW Overview ...

348

EM Waste Processing Advanced Joule-Heated November 17, 2010  

Hanford High Chromium/Sulfur HLW Elevated temperature, Enhanced formulation, Optimized bubbling Two more DM100 melter tests are scheduled for early December.

349

app_d  

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

DOEEIS-0287 Idaho HLW & FD EIS - New Information - Document 52, Hanford Advisory Board (Merilyn Reeves), Richland, WA Page 1 of 3 Document 52, Hanford Advisory Board (Merilyn...

350

Technology Development for Nuclear Waste Stabilization I  

Science Conference Proceedings (OSTI)

Oct 9, 2012 ... These systems have supported vitrification facilities at West Valley, DWPF and M- Area at Savannah River, WTP HLW and LAW at Hanford,...

351

Borosilicate Glass Formulations for Advanced Joule Heated Melters  

Summary Hanford High Cr/S HLW Selected formulations have waste loadings of 40 and 45 wt% ~23 38% increase over previous glass formulations

352

Borosilicate Glass Formulations for CCIM  

Hanford AZ-101 high-level waste (HLW) Hanford AN-105 low-activity waste (LAW) 3. Selection of Wastes Initially focused on Hanford waste streams

353

Nepheline Crystallization from Aluminosilicate Melts  

Al-limited and Al/Na-limited Hanford HLW. Figure 3. Crystallization during canister coolingis principal waste loading limiting factor . Current Nepheline Discriminator.

354

WTP: Challenges and Major Breakthroughs in High Level Waste ...  

Science Conference Proceedings (OSTI)

Abstract Scope, The US DOE has developed glass property-composition models to control glass compositions for HLW vitrification at Hanford Waste Treatment...

355

Challenges of Nuclear Waste Vitrification  

Science Conference Proceedings (OSTI)

The US DOE has developed glass property-composition models to control glass compositions for HLW vitrification at Hanford Waste Treatment & Immobilization...

356

Pulmonary macrophage and epithelial cells  

SciTech Connect

Separate abstracts were prepared for the 41 papers presented at the conference. Abstracts of two papers have appeared in previous issues of Energy Research Abstracts. (HLW)

Sanders, C.L.; Schneider, R.P.; Dagle, G.E.; Ragan, H.A. (eds.)

1977-01-01T23:59:59.000Z

357

Support for the in situ vitrification treatability study at the Idaho National Engineering Laboratory: FY 1988 summary  

SciTech Connect

The objective of this project is to determine if in situ vitrification (ISV) is a viable, long-term confinement technology for previously buried solid transuranic and mixed waste at the Radioactive Waste Management Complex (RWMC). The RWMC is located at the Idaho National Engineering Laboratory (INEL). In situ vitrification is a thermal treatment process that converts contaminated soils and wastes into a durable glass and crystalline form. During processing, heavy metals or other inorganic constituents are retained and immobilized in the glass structure, and organic constituents are typically destroyed or removed for capture by an off-gas treatment system. The primary FY 1988 activities included engineering-scale feasibility tests on INEL soils containing a high metals loading. Results of engineering-scale testing indicate that wastes with a high metals content can be successfully processed by ISV. The process successfully vitrified soils containing localized metal concentrations as high as 42 wt % without requiring special methods to prevent electrical shorting within the melt zone. Vitrification of this localized concentration resulted in a 15.9 wt % metals content in the entire ISV test block. This ISV metals limit is related to the quantity of metal that accumulates at the bottom of the molten glass zone. Intermediate pilot-scale testing is recommended to determine metals content scale-up parameters in order to project metals content limits for large-scale ISV operation at INEL.

Oma, K.H.; Reimus, M.A.H.; Timmerman, C.L.

1989-02-01T23:59:59.000Z

358

TSA waste stream and final waste form composition  

SciTech Connect

A final vitrified waste form composition, based upon the chemical compositions of the input waste streams, is recommended for the transuranic-contaminated waste stored at the Transuranic Storage Area of the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The quantities of waste are large with a considerable uncertainty in the distribution of various waste materials. It is therefore impractical to mix the input waste streams into an ``average`` transuranic-contaminated waste. As a result, waste stream input to a melter could vary widely in composition, with the potential of affecting the composition and properties of the final waste form. This work examines the extent of the variation in the input waste streams, as well as the final waste form under conditions of adding different amounts of soil. Five prominent Rocky Flats Plant 740 waste streams are considered, as well as nonspecial metals and the ``average`` transuranic-contaminated waste streams. The metals waste stream is the most extreme variation and results indicate that if an average of approximately 60 wt% of the mixture is soil, the final waste form will be predominantly silica, alumina, alkaline earth oxides, and iron oxide. This composition will have consistent properties in the final waste form, including high leach resistance, irrespective of the variation in waste stream. For other waste streams, much less or no soil could be required to yield a leach resistant waste form but with varying properties.

Grandy, J.D.; Eddy, T.L.; Anderson, G.L.

1993-01-01T23:59:59.000Z

359

UTILIZING STATISTICS TO DETERMINE HOW MUCH SAMPLING AND ANALYSISIS WARRANTED TO SUPPORT SAVANNAH RIVER SITEHIGH LEVEL WASTE SLUDGE BATCH PREPARATION  

SciTech Connect

Accelerated cleanup initiatives at the SRS include expediting radioactive sludge processing. Sludge is the highest risk component of waste since it contains the highest concentrations of long-lived radionuclides. The sludge is staged into ''batches'' that are then the feed material to the Defense Waste Processing Facility (DWPF) which vitrifies the waste into a safe form for permanent disposal. The preparation of each batch includes sampling and analysis of the slurried material. The results of the characterization are used as the bases for batch blending and processing decisions. Uncertainty is inherent in the information used for planning. There is uncertainty in the quantity of sludge contained in a tank, the waste composition, and the waste physical properties. The goal of this analysis is to develop the basis for the number of physical samples that should be taken from the slurried waste tank and the number of replicates of laboratory measurements that should be performed in order to achieve a specified uncertainty level. Recommendations for sampling and analysis strategies are made based on the results of the analysis.

Hamm, B

2007-05-17T23:59:59.000Z

360

An Investigation into the Oxidation State of Molybdenum in Simplified High Level Nuclear Waste Glass Compositions  

E-Print Network (OSTI)

a full simulated HLW stream based upon 4:1 ratio of high burn up UO2/mixed oxide (HBU/MOX) fuel. EXPERIMENTAL A series of simplified simulated HLW glasses (based on the 4:1 HBU/MOX composition) were melted

Sheffield, University of

Note: This page contains sample records for the topic "vitrified hlw quantities" 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

Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)  

SciTech Connect

The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

Burgard, K.C.

1998-04-09T23:59:59.000Z

362

Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)  

SciTech Connect

The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

Burgard, K.C.

1998-06-02T23:59:59.000Z

363

Microsoft PowerPoint - 2-04_Vienna Glass Models Draft (11-15-10) v2.pptx  

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

D. J.D. Vienna D. J.D. Vienna Vienna Pacific Northwest National Laboratory, Richland, WA J.D. Belsher, P.A. Empey, and F.L. Meinert J.D. Belsher, P.A. Empey, and F.L. Meinert Washington River Protection Solutions, Richland, WA Property Models for High Waste Property Models for High Waste Loaded Hanford HLW Glasses Loaded Hanford HLW Glasses Print Close 2  Waste loading improvements in Hanford HLW glasses  Glass property models  Systems studies with dynamic flowsheet model  waste glass predictions for Hanford  influential components  glass testing priorities  Status of the testing/modeling task  Summary and conclusions  Acknowledgements Outline Print Close Waste Loading Improvements in Hanford HLW Glass  Hanford HLW covers a broad range of compositions -

364

Systems study of the feasibility of high-level nuclear waste fractionation for thermal stress control in a geologic repository: appendices  

Science Conference Proceedings (OSTI)

This study assesses the benefits and costs of fractionating the cesium and strontium (Cs/Sr) components in commercial high-level waste (HLW) to a separate waste stream for the purpose of reducing geologic-repository thermal stresses in the region of the HLW. The major conclusion is that the Cs/Sr fractionation concept offers the prospect of a substantial total system cost advantage for HLW disposal if reduced HLW package temperatures in a basalt repository are desired. However there is no cost advantage if currently designated maximum design temperatures are acceptable. Aging the HLW for 50 to 100 years can accomplish similar results at equivalent or lower costs. Volume II contains appendices for: (1) thermal analysis supplement; (2) fractionation process experimental results supplement; (3) cost analysis supplement; and (4) radiological risk analysis supplement.

McKee, R.W.; Elder, H.K.; McCallum, R.F.; Silviera, D.J.; Swanson, J.L.; Wiles, L.E.

1983-06-01T23:59:59.000Z

365

Hanford waste tanks - light at the end of the tunnel  

DOE Green Energy (OSTI)

The U.S. Department of Energy (DOE) faced several problems in its Hanford Site tank farms in the early nineties. It had 177 waste tanks, ranging in size from 55,000 to 1,100,000 gallons, which contained more than 55 million gallons of liquid and solid high-level radioactive waste (HLW) from a variety of processes. Unfortunately, waste transfer records were incomplete. Chemical reactions going on in the tanks were not totally understood. Every tank had high concentrations of powerful oxidizers in the form of nitrates and nitrites, and some tanks had relatively high concentrations of potential fuels that could react explosively with oxidizers. A few of these tanks periodically released large quantities of hydrogen and nitrous oxide, a mixture that was potentially more explosive than hydrogen and air. Both the nitrate/fuel and hydrogen/nitrous oxide reactions had the potential to rupture a tank exposing workers and the general public to unacceptably large quantities of radioactive material. One tank (241-C-106) was generating so much heat that water had to be added regularly to avoid thermal damage to the tank's concrete exterior shell. The tanks contained more than 250 million Curies of radioactivity. Some of that radioactivity was in the form of fissile plutonium, which represented a potential criticality problem. As awareness of the potential hazards grew, the public and various regulatory agencies brought increasing pressure on DOE to quantify the hazards and mitigate any that were found to be outside accepted risk guidelines. In 1990, then Representative, now Senator Ron Wyden (D-Oregon), introduced an amendment to Public Law 101-510, Section 3137, that required DOE to identify Hanford tanks that might have a serious potential for release of high-level waste.

POPPITI, J.A.

1999-09-29T23:59:59.000Z

366

STATISTICAL EVALUATION OF SMALL SCALE MIXING DEMONSTRATION SAMPLING AND BATCH TRANSFER PERFORMANCE - 12093  

SciTech Connect

The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS) has previously presented the results of mixing performance in two different sizes of small scale DSTs to support scale up estimates of full scale DST mixing performance. Currently, sufficient sampling of DSTs is one of the largest programmatic risks that could prevent timely delivery of high level waste to the WTP. WRPS has performed small scale mixing and sampling demonstrations to study the ability to sufficiently sample the tanks. The statistical evaluation of the demonstration results which lead to the conclusion that the two scales of small DST are behaving similarly and that full scale performance is predictable will be presented. This work is essential to reduce the risk of requiring a new dedicated feed sampling facility and will guide future optimization work to ensure the waste feed delivery mission will be accomplished successfully. This paper will focus on the analytical data collected from mixing, sampling, and batch transfer testing from the small scale mixing demonstration tanks and how those data are being interpreted to begin to understand the relationship between samples taken prior to transfer and samples from the subsequent batches transferred. An overview of the types of data collected and examples of typical raw data will be provided. The paper will then discuss the processing and manipulation of the data which is necessary to begin evaluating sampling and batch transfer performance. This discussion will also include the evaluation of the analytical measurement capability with regard to the simulant material used in the demonstration tests. The paper will conclude with a discussion of the analysis results illustrating the relationship between the pre-transfer samples and the batch transfers, which support the recommendation regarding the need for a dedicated feed sampling facility.

GREER DA; THIEN MG

2012-01-12T23:59:59.000Z

367

Economic Feasibility of Electrochemical Caustic Recycling at the Hanford Site  

SciTech Connect

This report contains a review of potential cost benefits of NaSICON Ceramic membranes for the separation of sodium from Hanford tank waste. The primary application is for caustic recycle to the Waste Treatment and Immobilization Plant (WTP) pretreatment leaching operation. The report includes a description of the waste, the benefits and costs for a caustic-recycle facility, and Monte Carlo results obtained from a model of these costs and benefits. The use of existing cost information has been limited to publicly available sources. This study is intended to be an initial evaluation of the economic feasibility of a caustic recycle facility based on NaSICON technology. The current pretreatment flowsheet indicates that approximately 6,500 metric tons (MT) of Na will be added to the tank waste, primarily for removing Al from the high-level waste (HLW) sludge (Kirkbride et al. 2007). An assessment (Alexander et al. 2004) of the pretreatment flowsheet, equilibrium chemistry, and laboratory results indicates that the quantity of Na required for sludge leaching will increase by 6,000 to 12,000 MT in order to dissolve sufficient Al from the tank-waste sludge material to maintain the number of HLW canisters produced at 9,400 canisters as defined in the Office of River Protection (ORP) System Plan (Certa 2003). This additional Na will significantly increase the volume of LAW glass and extend the processing time of the Waste Treatment and Immobilization Plant (WTP). Future estimates on sodium requirements for caustic leaching are expected to significantly exceed the 12,000-MT value and approach 40,000-MT of total sodium addition for leaching (Gilbert, 2007). The cost benefit for caustic recycling is assumed to consist of four major contributions: 1) the cost savings realized by not producing additional immobilized low-activity waste (ILAW) glass, 2) caustic recycle capital investment, 3) caustic recycle operating and maintenance costs, and 4) research and technology costs needed to deploy the technology. In estimating costs for each of these components, several parameters are used as inputs. Due to uncertainty in assuming a singular value for each of these parameters, a range of possible values is assumed. A Monte Carlo simulation is then performed where the range of these parameters is exercised, and the resulting range of cost benefits is determined.

Poloski, Adam P.; Kurath, Dean E.; Holton, Langdon K.; Sevigny, Gary J.; Fountain, Matthew S.

2009-03-01T23:59:59.000Z

368

Minor actinide waste disposal in deep geological boreholes  

E-Print Network (OSTI)

The purpose of this investigation was to evaluate a waste canister design suitable for the disposal of vitrified minor actinide waste in deep geological boreholes using conventional oil/gas/geothermal drilling technology. ...

Sizer, Calvin Gregory

2006-01-01T23:59:59.000Z

369

The Hanford waste feed delivery operational research model  

Science Conference Proceedings (OSTI)

The Hanford cleanup mission is to vitrify 56 million gallons of nuclear waste, currently stored in 177 underground tanks, at the Waste Treatment and Immobilization Plant (WTP). The WTP operations begin in 2019. Waste transfers from the Tank Farms to ...

Joanne Berry; Vishvas Patel; Karthik Vasudevan

2011-12-01T23:59:59.000Z

370

Sludge Mass Reduction Update  

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

Program Status EM-21 Technical Exchange Neil Davis May 19, 2009 LWO-SPT-2009-00079 2 LWO-SPT-2009-00079 SRS LWO Mission Receive waste Store waste safely Vitrify waste Process salt...

371

RADIOACTIVE DEMONSTRATIONS OF FLUIDIZED BED STEAM REFORMING AS A SUPPLEMENTARY TREATMENT FOR HANFORD'S LOW ACTIVITY WASTE AND SECONDARY WASTES  

SciTech Connect

The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as Cs-137, I-129, Tc-99, Cl, F, and SO4 that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750 C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the Savannah River National Laboratory (SRNL) to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of I-125/129 and Tc-99 to chemically resemble WTP-SW. Ninety six grams of radioactive product were made for testing. The second campaign commenced using SRS LAW chemically trimmed to look like Hanford's LAW. Six hundred grams of radioactive product were made for extensive testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.

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

2011-02-24T23:59:59.000Z

372

RADIOACTIVE DEMONSTRATION OF FINAL MINERALIZED WASTE FORMS FOR HANFORD WASTE TREATMENT PLANT SECONDARY WASTE BY FLUIDIZED BED STEAM REFORMING USING THE BENCH SCALE REFORMER PLATFORM  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as {sup 137}Cs, {sup 129}I, {sup 99}Tc, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750 C) continuous method by which WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage, but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the SRNL to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. BSR testing with WTP SW waste surrogates and associated analytical analyses and tests of granular products (GP) and monoliths began in the Fall of 2009, and then was continued from the Fall of 2010 through the Spring of 2011. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of {sup 125/129}I and {sup 99}Tc to chemically resemble WTP-SW. Prior to these radioactive feed tests, non-radioactive simulants were also processed. Ninety six grams of radioactive granular product were made for testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.

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

2012-02-02T23:59:59.000Z

373

acre feet (af) -A quantity of volume of water that covers one acre to a depth of one foot; equal to 43,560 cubic feet or 325,851 gallons.  

E-Print Network (OSTI)

it is not hydrostatically connected. per capita water use - Water produced by or introduced into the system of a water of a fluid into, through, or from a porous medium. self-produced water - Water supply (usually from wells) developed and used by an individual or entity. Also called self-produced water. self-supplied water - Water

Lund, Jay R.

374

Activities on the ORR have the potential to release small quantities of radionuclides and hazardous chemicals to the environment. These releases could result in exposures of members of the public to low  

E-Print Network (OSTI)

Biological Effects Quality Assurance in Monitoring programme BFR Brominated flame retardant BNFL British Removal Plant (at BNFL Sellafield) EDCs Endocrine Disrupting Chemicals EDMAR Endocrine Disruption Similarity Percentages Routine SIXEP Site Ion Exchange Effluent Plant (at BNFL Sellafield) SPI Sediment

Pennycook, Steve

375

Activities on the ORR have the potential to release small quantities of radionuclides and hazardous chemicals to the environment. These releases could result in exposures of members of the public to low  

E-Print Network (OSTI)

Bromwich, T. (EPSRC DTA) Brown, G. (CASE: Colebrand Ltd.) Campbell, P.J.D. (CASE: BNFL) Carter, R. (EPSRC Australian Government BNFL BAE British Council British National Space Agency Chinese Government Clarendon

Pennycook, Steve

376

Activities on the Oak Ridge Reservation have the potential to release small quantities of radionuclides and hazardous chemicals to the environment. These releases could result in exposures of members of the  

E-Print Network (OSTI)

SINCE 1997 3.1. International Atomic Energy Agency (IAEA) 3.2. BNFL and Pangea 3.3. British Geological and practice since 1997 is reviewed. The 1994 IAEA guidelines are being replaced by a new IAEA draft. BNFL held Autorité de Sûreté Nucléaire (French Nuclear Safety Authority) BGS British Geological Survey BNFL British

Pennycook, Steve

377

Genital Herpes Evaluation by Quantitative TaqMan PCR: Correlating Single Detection and Quantity of HSV-2 DNA in Cervicovaginal Lavage Fluids with Cross-sectional and Longitudinal Clinical Data.  

E-Print Network (OSTI)

by quantitative TaqMan PCR: correlating single detection andof a quantitative competitive PCR assay for measuring herpessamples by a real-time taqman PCR assay. J Med Virol 2005,

2010-01-01T23:59:59.000Z

378

Accident analysis for high-level waste management alternatives in the US Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement  

SciTech Connect

A comparative generic accident analysis was performed for the programmatic alternatives for high-level waste (HLW) management in the US Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement (EM PEIS). The key facilities and operations of the five major HLW management phases were considered: current storage, retrieval, pretreatment, treatment, and interim canister storage. A spectrum of accidents covering the risk-dominant accidents was analyzed. Preliminary results are presented for HLW management at the Hanford site. A comparison of these results with those previously advanced shows fair agreement.

Folga, S.; Mueller, C.; Roglans-Ribas, J.

1994-02-01T23:59:59.000Z

379

RIVER PROTECTION PROJECT MISSION ANALYSIS WASTE BLENDING STUDY  

SciTech Connect

Preliminary evaluation for blending Hanford site waste with the objective of minimizing the amount of high-level waste (HLW) glass volumes without major changes to the overall waste retrieval and processing sequences currently planned. The evaluation utilizes simplified spreadsheet models developed to allow screening type comparisons of blending options without the need to use the Hanford Tank Waste Operations Simulator (HTWOS) model. The blending scenarios evaluated are expected to increase tank farm operation costs due to increased waste transfers. Benefit would be derived from shorter operating time period for tank waste processing facilities, reduced onsite storage of immobilized HLW, and reduced offsite transportation and disposal costs for the immobilized HLW.

SHUFORD DH; STEGEN G

2010-04-19T23:59:59.000Z

380

Bases, Assumptions, and Results of the Flowsheet Calculations for the Decision Phase Salt Disposition Alternatives  

SciTech Connect

The High Level Waste (HLW) Salt Disposition Systems Engineering Team was formed on March 13, 1998, and chartered to identify options, evaluate alternatives, and recommend a selected alternative(s) for processing HLW salt to a permitted wasteform. This requirement arises because the existing In-Tank Precipitation process at the Savannah River Site, as currently configured, cannot simultaneously meet the HLW production and Authorization Basis safety requirements. This engineering study was performed in four phases. This document provides the technical bases, assumptions, and results of this engineering study.

Dimenna, R.A.; Jacobs, R.A.; Taylor, G.A.; Durate, O.E.; Paul, P.K.; Elder, H.H.; Pike, J.A.; Fowler, J.R.; Rutland, P.L.; Gregory, M.V.; Smith III, F.G.; Hang, T.; Subosits, S.G.; Campbell, S.G.

2001-03-26T23:59:59.000Z

Note: This page contains sample records for the topic "vitrified hlw quantities" 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.


381

EIS-0287: Notice of Preferred Sodium Bearing Waste Treatment Technology |  

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

Preferred Sodium Bearing Waste Treatment Preferred Sodium Bearing Waste Treatment Technology EIS-0287: Notice of Preferred Sodium Bearing Waste Treatment Technology Idaho High-Level Waste (HLW) and Facilities Disposition In October 2002, the U.S. Department of Energy (DOE or the Department) issued the Final Idaho High-Level Waste (HLW) and Facilities Disposition Environmental Impact Statement (DOE/EIS-0287 (Final EIS)). The Final EIS contains an evaluation of reasonable alternatives for the management of mixed transuranic waste/sodium bearing waste (SBW),1 mixed HLW calcine, and associated low-level waste (LLW), as well as disposition alternatives for HLW facilities when their missions are completed. DOE/EIS-0287, Notice of Preferred Sodium Bearing Waste Treatment Technology, Office of Environmental Management, Idaho, 70 FR 44598 (August

382

ch_5  

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

25 25 DOE/EIS-0287 Idaho HLW & FD EIS 5.3 Facility Disposition Impacts Section 5.3 presents a discussion of potential impacts associated with the disposition of exist- ing HLW management facilities at INEEL and disposition of new facilities that would be built in support of the proposed waste processing alternatives. The discussion includes (1) the potential impacts of short-term actions in dispo- sitioning new and existing HLW management facilities, (2) the potential long-term impacts from the disposal of the grouted low-level waste fraction in either a new disposal facility at INTEC or in the Tank Farm and bin sets, and (3) the potential long-term impacts of residual con- tamination in closed HLW management facili- ties. The six facility disposition alternatives are

383

High Level Waste Corporate Board Charter  

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

on 24 July 2008 1 on 24 July 2008 1 Office of Environmental Management High-Level Waste Corporate Board Charter Purpose This Charter establishes the High- Level Waste (HLW) Corporate Board, (hereinafter referred to as the 'Board') within the Office of Environmental Management (EM). The Board will serve as a consensus building body to integrate the Department of Energy (DOE) HLW management and disposition activities across the EM program and, with the coordination and cooperation of other program offices, across the DOE complex. The Board will identify the need for and develop policies, planning, standards and guidance and provide the integration necessary to implement an effective and efficient national HLW program. The Board will also evaluate the implications of HLW issues and their

384

Review of the Hanford Site Waste Treatment and Immobilization...  

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

U.S. Department of Energy DOE-WTP ORP WTP Project Office HLW High-Level Waste Facility HVAC Heating, Ventilation, and Air Conditioning LAB Analytical Laboratory LAW Low-Activity...

385

Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report  

Science Conference Proceedings (OSTI)

SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described.

SK Sundaram; ML Elliott; D Bickford

1999-11-19T23:59:59.000Z

386

Microsoft PowerPoint - IPRC-2012_Kormilitsyn_RAW+ [Compatibility...  

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

reactors Fast Reactors Model of RUSSIANIntegrated UNF Management System for 2025-2030 RBMK 7 units VVER- 10001200 30 units (in Russia) U fuel fab HLW Disposal Fast Reactor...

387

Microsoft Word - EM SSAB Chairs Meeting Oct 2012 Minutes_2_14...  

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

section 435.1enables EM to classify some equipment and wastes used or produced in HLW treatment programs as LLW or transuranic waste-based on its actual characteristics, not...

388

PRESENT STATUS OF THE OMEGA PROGRAM IN JAPAN PREPARED FOR SECOND GENEWIL MEETING FOR INFORiVIATION EXCANGE MEETING ON ACTINIDE AND FISSION PRODUCT SEPARATION AND TRANSMUTATION  

E-Print Network (OSTI)

The management of high. level radioactive waste ( HLW) generated from the reprocessing of spent fuel is very important as well as the safety assurance, to further develop nuclear electricity generation.

Hiroyuki Yoshiiia; Satoshi Tani; Tadashi Inoue

1992-01-01T23:59:59.000Z

389

Pacific Northwest Laboratory annual report for 1978 to the DOE Assistant Secretary for Environment. Part 1. Biomedical sciences  

SciTech Connect

Separate abstracts were prepared for the 80 papers of the report. Tabulated data on dose-effect studies with inhaled Pu in beagles are given in an appendix. (HLW)

Wiley, W.R.

1979-02-01T23:59:59.000Z

390

Proceedings of thermal ecology II  

SciTech Connect

Separate abstracts were prepared for fifty-one papers presented at the conference. An additional seven papers were presented for which abstracts appeared in previous issues of ERA. (HLW)

Esch, G.W.; McFarlane, R.W. (eds.)

1976-01-01T23:59:59.000Z

391

EIS-0220: Supplemental Record of Decision | Department of Energy  

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

impacts of the newly-selected alternative were analyzed in the IMNM EIS. This alternative includes the transfer of the solutions to the SRS high-level waste (HLW)...

392

Tank Deployment Plan Overview for Next Generation Melter at WTP  

Primary NGM Decisions (DOE-EM R&D Plan) Time Frame Select NGM Test Platforms for R&D 2011 Down-Select NGM Melter Technologies 2013/14 Select HLW and LAW NGM

393

--No Title--  

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

will conduct a mixing demonstration of the Hanford AY-102 High level Waste (HLW) tank. This testing will be conducted in 786-A. The test is a visual type of demonstration where no...

394

app_d  

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

DOEEIS-0287 Idaho HLW & FD EIS Document 1, Darryl D. Siemer, Idaho Falls, ID Page 1 of 18 Document 1, Darryl D. Siemer, Idaho Falls, ID Page 2 of 18 - New Information - DOE...

395

app_d  

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

7 DOEEIS-0287 Idaho HLW & FD EIS - New Information - Document 54, INEEL Citizens Advisory Board (Stan Hobson), Idaho Falls, ID Page 1 of 9 Document 53, Public Comment Hearing,...

396

Summary.qxd  

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

3 DOEEIS-0287 Idaho HLW & FD EIS 6.4.4 HEALTH AND SAFETY Airborne contamination is the principal transport pathway through which radioactive materials from the INEEL affect...

397

Summary.qxd  

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

9 DOEEIS-0287 Idaho HLW & FD EIS 5.0 Areas of Controversy There are areas relevant to alternatives consid- ered in this EIS, where viewpoints may differ among members of the...

398

Summary.qxd  

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

19 DOEEIS-0287 Idaho HLW & FD EIS nition of a Class A low-level waste. Under the Planning Basis Option, DOE would dis- pose of the Class A-type grout in an offsite low-level waste...

399

ch_3  

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

13 DOEEIS-0287 Idaho HLW & FD EIS except the pillar and panel tanks) would be full of mixed transuranic waste in approximately 2017. Other facilities depending on the capacity of...

400

Summary.qxd  

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

DOEEIS-0287 Idaho HLW & FD EIS 2.0 Activities since the Issuance of the Draft EIS 2.1 Summary of Public Comments and Agency Responses The Draft EIS was mailed to the public and...

Note: This page contains sample records for the topic "vitrified hlw quantities" 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.


401

app_d  

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

5 DOEEIS-0287 Idaho HLW & FD EIS Document 1, Darryl D. Siemer, Idaho Falls, ID Page 9 of 18 Document 1, Darryl D. Siemer, Idaho Falls, ID Page 10 of 18 - New Information - DOE...

402

app_d  

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

Boise, ID Page 2 of 2 Document 66, U.S. EPA-Region 10 (Christian F. Gebhardt), Seattle, WA Page 1 of 1 D-171 DOEEIS-0287 Idaho HLW & FD EIS - New Information - Document...

403

Applied health physics and safety annual report for 1976  

SciTech Connect

Progress is reported in the following areas of research: personnel monitoring; health physics instrumentation; atmospheric monitoring; water monitoring; radiation background measurements; soil samples; laboratory operations monitoring; radiation incidents; laundry monitoring; accident analysis; and industrial safety. (HLW)

Auxier, J.A.; Davis, D.M.

1977-08-01T23:59:59.000Z

404

app_d  

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

199 DOEEIS-0287 Idaho HLW & FD EIS - New Information - Document 80, Melissa Clark Rhodes, Jackson, WY Page 11 of 19 Document 80, Melissa Clark Rhodes, Jackson, WY Page 12 of 19...

405

app_d  

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

2 of 2 - New Information - D-21 DOEEIS-0287 Idaho HLW & FD EIS Document 14, Melissa Clark Rhodes, Jackson, WY Page 1 of 2 Document 14, Melissa Clark Rhodes, Jackson, WY Page 2...

406

app_d  

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

203 DOEEIS-0287 Idaho HLW & FD EIS - New Information - Document 80, Melissa Clark Rhodes, Jackson, WY Page 19 of 19 Document 81, Dennis Donnelly, Pocatello, ID Page 1 of 2 DOE...

407

2011 Meeting  

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

HLW, support for bilateral interactions between the United States and the Republic of Korea and the United States and Japan, and planning for U.S. involvement in FY 2012 in...

408

Slide 1  

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

A.D. Cozzi and J.D. Newell Technical Need * Thermal treatment systems planned for the Hanford Waste Treatment Plant (WTP) immobilization of High-Level Waste (HLW), Low Activity...

409

Fate of Tc99 at WTP and Current Work on Capture  

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

1 Technetium DF (per Rhenium simulant) from VSL-04R4710-1, Melter DM-100 Tests (1) 6 A. Retention of Tc 99 in LAW and HLW Glass (cont.) Table 2 Technetium DF (per Rhenium...

410

EPRI Review of Geologic Disposal for Used Fuel and High Level Radioactive Waste: Volume IV - Lessons Learned  

Science Conference Proceedings (OSTI)

The effective termination of the Yucca Mountain program by the U.S. Administration in 2009 has further delayed the construction and operation of a permanent disposal facility for used fuel and high level radioactive waste (HLW) in the United States. In concert with this decision, the President directed the Energy Secretary to establish the Blue Ribbon Commission on America's Nuclear Future to review and provide recommendations on options for managing used fuel and HLW. EPRI is uniquely positioned to prov...

2010-09-29T23:59:59.000Z

411

Preliminary waste form characteristics report Version 1.0. Revision 1  

SciTech Connect

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

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

1991-10-11T23:59:59.000Z

412

SLUDGE HEEL REMOVAL BY ALUMINUM DISSOLUTION AT SAVANNAH RIVER SITE 12390  

SciTech Connect

High Level Waste (HLW) at the Savannah River Site (SRS) is currently stored in aging underground storage tanks. This waste is a complex mixture of insoluble solids, referred to as sludge, and soluble salts. Continued long-term storage of these radioactive wastes poses an environmental risk. Operations are underway to remove and disposition the waste, clean the tanks and fill with grout for permanent closure. Heel removal is the intermediate phase of the waste retrieval and tank cleaning process at SRS, which is intended to reduce the volume of waste prior to treatment with oxalic acid. The goal of heel removal is to reduce the residual amount of radioactive sludge wastes to less than 37,900 liters (10,000 gallons) of wet solids. Reducing the quantity of residual waste solids in the tank prior to acid cleaning reduces the amount of acid required and reduces the amount of excess acid that could impact ongoing waste management processes. Mechanical heel removal campaigns in Tank 12 have relied solely on the use of mixing pumps that have not been effective at reducing the volume of remaining solids. The remaining waste in Tank 12 is known to have a high aluminum concentration. Aluminum dissolution by caustic leaching was identified as a treatment step to reduce the volume of remaining solids and prepare the tank for acid cleaning. Dissolution was performed in Tank 12 over a two month period in July and August, 2011. Sample results indicated that 16,440 kg of aluminum oxide (boehmite) had been dissolved representing 60% of the starting inventory. The evolution resulted in reducing the sludge solids volume by 22,300 liters (5900 gallons), preparing the tank for chemical cleaning with oxalic acid.

Keefer, M.

2012-01-12T23:59:59.000Z

413

Preliminary estimates of the total-system cost for the restructured program: An addendum to the May 1989 analysis of the total-system life cycle cost for the Civilian Radioactive Waste Management Program  

SciTech Connect

The total-system life-cycle cost (TSLCC) analysis for the Department of Energy`s (DOE) Civilian Radioactive Waste Management Program is an ongoing activity that helps determine whether the revenue-producing mechanism established by the Nuclear Waste Policy Act of 1982 - a fee levied on electricity generated and sold by commercial nuclear power plants - is sufficient to cover the cost of the program. This report provides cost estimates for the sixth annual evaluation of the adequacy of the fee. The costs contained in this report represent a preliminary analysis of the cost impacts associated with the Secretary of Energy`s Report to Congress on Reassessment of the Civilian Radioactive Waste Management Program issued in November 1989. The major elements of the restructured program announced in this report which pertain to the program`s life-cycle costs are: a prioritization of the scientific investigations program at the Yucca Mountain candidate site to focus on identification of potentially adverse conditions, a delay in the start of repository operations until 2010, the start of limited waste acceptance at the monitored retrievable storage (MRS) facility in 1998, and the start of waste acceptance at the full-capability MRS facility in 2,000. Based on the restructured program, the total-system cost for the system with a repository at the candidate site at Yucca Mountain in Nevada, a facility for monitored retrievable storage (MRS), and a transportation system is estimated at $26 billion (expressed in constant 1988 dollars). In the event that a second repository is required and is authorized by the Congress, the total-system cost is estimated at $34 to $35 billion, depending on the quantity of spent fuel and high-level waste (HLW) requiring disposal. 17 figs., 17 tabs.

NONE

1990-12-01T23:59:59.000Z

414

Pretreatment of neutralized cladding removal waste sludge: Results of the second design basis experiment  

SciTech Connect

For several years, the Pacific Northwest Laboratory (PNL) has been investigating methods to pretreat Hanford neutralized cladding removal waste (NCRW) sludge. In the past, Zircaloy-clad metallic U fuel was chemically decladded using the Zirflex process; NCRW sludge was formed when the decladding solution was neutralized for storage in carbon-steel tanks. This sludge, which is currently stored in Tanks 103-AW and 105-AW on the Hanford Site, primarily consists of insoluble Zr hydroxides and/or oxides and NaF. Significant quantities of Al, La, U, as well as other insoluble minor constituents are present in the sludge, along with sodium and potassium nitrates, nitrites, and hydroxides in the interstitial liquid. The sludge contains about 2,000 nCi of transuranic (TRU) material per gram of dry sludge, and mixed fission products. Therefore, the sludge must be handled as high-level waste (HLW). The NCRW sludge must be pretreated before treatment (e.g., vitrification) and disposal, so that the overall cost of disposal can be minimized. The NCRW pretreatment flowsheet was designed to achieve the following objectives: (a) to separate Am and Pu from the major sludge constituents (Na, Zr). (b) to separate Am and Pu from U. (c) to concentrate Am and Pu in a small volume for immobilization in borosilicate glass, based on Hanford Waste Vitrification Plant (HWVP). The flowsheet involves: (1) sludge washing, (2) sludge dissolution, (3) extraction of U with tributyl phosphate (TBP), and (4) extraction of TRUs with octyl(phenyl)-N,N-diisobutlycarbamoylmethyl-phosphine oxide (CMPO). As presented in the flowsheet, the NCRW sludge is first washed with 0.I M NaOH to remove interstitial liquid and soluble salts from the sludge including sodium and potassium fluorides, carbonates, hydroxides, nitrates, and nitrites. The washed sludge is then subjected to two dissolution steps to achieve near complete dissolution of Zr.

Lumetta, G.J.

1994-05-01T23:59:59.000Z

415

Comparison of cask and drywell storage concepts for a monitored retrievable storage/interim storage system  

SciTech Connect

The Department of Energy, through its Richland Operations Office is evaluating the feasibility, timing, and cost of providing a federal capability for storing the spent fuel, high-level wastes, and transuranic wastes that DOE may be obligated by law to manage until permanent waste disposal facilities are available. Three concepts utilizing a monitored retrievable storage/interim storage (MRS/IS) facility have been developed and analyzed. The first concept, co-location with a reprocessing plant, has been developed by staff of Allied General Nuclear Services. the second concept, a stand-alone facility, has been developed by staff of the General Atomic Company. The third concept, co-location with a deep geologic repository, has been developed by the Pacific Northwest Laboratory with the assistance of the Westinghouse Hanford Company and Kaiser Engineers. The objectives of this study are: to develop preconceptual designs for MRS/IS facilities: to examine various issues such as transportation of wastes, licensing of the facilities, and environmental concerns associated with operation of such facilities; and to estimate the life-cycle costs of the facilities when operated in response to a set of scenarios that define the quantities and types of waste requiring storage in specific time periods, generally spanning the years 1989 to 2037. Three scenarios are examined to develop estimates of life-cycle costs for the MRS/IS facilities. In the first scenario, the reprocessing plant is placed in service in 1989 and HLW canisters are stored until a repository is opened in the year 1998. Additional reprocessing plants and repositories are placed in service at intervals as needed to meet the demand. In the second scenario, the reprocessing plants are delayed in starting operations by 10 years, but the repositories open on schedule. In the third scenario, the repositories are delayed 10 years, but the reprocessing plants open on schedule.

Rasmussen, D.E.

1982-12-01T23:59:59.000Z

416

Development and application of a conceptual approach for defining high-level waste  

SciTech Connect

This paper presents a conceptual approach to defining high-level radioactive waste (HLW) and a preliminary quantitative definition obtained from an example implementation of the conceptual approach. On the basis of the description of HLW in the Nuclear Waste Policy Act of 1982, we have developed a conceptual model in which HLW has two attributes: HLW is (1) highly radioactive and (2) requires permanent isolation via deep geologic disposal. This conceptual model results in a two-dimensional waste categorization system in which one axis, related to ''requires permanent isolation,'' is associated with long-term risks from waste disposal and the other axis, related to ''highly radioactive,'' is associated with short-term risks from waste management and operations; this system also leads to the specification of categories of wastes that are not HLW. Implementation of the conceptual model for defining HLW was based primarily on health and safety considerations. Wastes requiring permanent isolation via deep geologic disposal were defined by estimating the maximum concentrations of radionuclides that would be acceptable for disposal using the next-best technology, i.e., greater confinement disposal (GCD) via intermediate-depth burial or engineered surface structures. Wastes that are highly radioactive were defined by adopting heat generation rate as the appropriate measure and examining levels of decay heat that necessitate special methods to control risks from operations in a variety of nuclear fuel-cycle situations. We determined that wastes having a power density >200 W/m/sup 3/ should be considered highly radioactive. Thus, in the example implementation, the combination of maximum concentrations of long-lived radionuclides that are acceptable for GCD and a power density of 200 W/m/sup 3/ provides boundaries for defining wastes that are HLW.

Croff, A.G.; Forsberg, C.W.; Kocher, D.C.; Cohen, J.J.; Smith, C.F.; Miller, D.E.

1986-01-01T23:59:59.000Z

417

Bioremediation of petroleum hydrocarbo-contaminated soils, comprehensive report, December 1999  

E-Print Network (OSTI)

completely degrade large quantities of oil. A demonstrationthe manufacture of motor oil, large quantities of sulfuric

Hazen, Terry

2000-01-01T23:59:59.000Z

418

Defense High Level Waste Disposal Container System Description  

Science Conference Proceedings (OSTI)

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

NONE

2000-10-12T23:59:59.000Z

419

LIFE EXTENSION PROGRAM FOR THE MODULAR CAUSTIC SIDE SOLVENT EXTRACTION UNIT AT SAVANNAH RIVER SITE  

Science Conference Proceedings (OSTI)

Caustic Side Solvent Extraction (CSSX) is currently used at the U.S. Department of Energy (DOE) Savannah River Site (SRS) for removal of cesium from the high-level salt-wastes stored in underground tanks. At SRS, the CSSX process is deployed in the Modular CSSX Unit (MCU). The CSSX technology utilizes a multi-component organic solvent and annular centrifugal contactors to extract cesium from alkaline salt waste. Coalescers and decanters process the Decontaminated Salt Solution (DSS) and Strip Effluent (SE) streams to allow recovery and reuse of the organic solvent and to limit the quantity of solvent transferred to the downstream facilities. MCU is operated in series with the Actinide Removal Process (ARP) which removes strontium and actinides from salt waste utilizing monosodium titanate. ARP and MCU were developed and implemented as interim salt processing until future processing technology, the CSSX-based Salt Waste Processing Facility (SWPF), is operational. SWPF is slated to come on-line in October 2014. The three year design life of the ARP/MCU process, however, was reached in April 2011. Nevertheless, most of the individual process components are capable of operating longer. An evaluation determined ARP/MCU can operate until 2015 before major equipment failure is expected. The three year design life of the ARP/MCU Life Extension (ARP/MCU LE) program will bridge the gap between current ARP/MCU operations and the start of SWPF operation. The ARP/MCU LE program introduces no new technologies. As a portion of this program, a Next Generation Solvent (NGS) and corresponding flowsheet are being developed to provide a major performance enhancement at MCU. This paper discusses all the modifications performed in the facility to support the ARP/MCU Life Extension. It will also discuss the next generation chemistry, including NGS and new stripping chemistry, which will increase cesium removal efficiency in MCU. Possible implementation of the NGS chemistry in MCU accomplishes two objectives. MCU serves as a demonstration facility for improved flowsheet deployment at SWPF; operating with NGS and boric acid validates improved cesium removal performance and increased throughput as well as confirms Defense Waste Processing Facility (DWPF) ability to vitrify waste streams containing boron. NGS implementation at MCU also aids the ARP/MCU LE operation, mitigating the impacts of delays and sustaining operations until other technology is able to come on-line.

Samadi-Dezfouli, A.

2012-11-14T23:59:59.000Z

420

TREATMENT OF GASEOUS EFFLUENTS ISSUED FROM RECYCLING A REVIEW OF THE CURRENT PRACTICES AND PROSPECTIVE IMPROVEMENTS  

Science Conference Proceedings (OSTI)

The objectives of gaseous waste management for the recycling of nuclear used fuel is to reduce by best practical means (ALARA) and below regulatory limits, the quantity of activity discharged to the environment. The industrial PUREX process recovers the fissile material U(VI) and Pu(IV) to re-use them for the fabrication of new fuel elements e.g. recycling plutonium as a Mixed Oxide (MOX) fuel or recycling uranium for new enrichment for Pressurized Water Reactor (PWR). Meanwhile the separation of the waste (activation and fission product) is performed as a function of their pollution in order to store and avoid any potential danger and release towards the biosphere. Raffinate, that remains after the extraction step and which contains mostly all fission products and minor actinides is vitrified, the glass package being stored temporarily at the recycling plant site. Hulls and end pieces coming from PWR recycled fuel are compacted by means of a press leading to a volume reduced to 1/5th of initial volume. An organic waste treatment step will recycle the solvent, mainly tri-butyl phosphate (TBP) and some of its hydrolysis and radiolytic degradation products such as dibutyl phosphate (HDPB) and monobutyl phosphate (H2MBP). Although most scientific and technological development work focused on high level waste streams, a considerable effort is still under way in the area of intermediate and low level waste management. Current industrial practices for the treatment of gaseous effluents focusing essentially on Iodine-129 and Krypton-85 will be reviewed along with the development of novel technologies to extract, condition, and store these fission products. As an example, the current industrial practice is to discharge Kr-85, a radioactive gas, entirely to the atmosphere after dilution, but for the large recycling facilities envisioned in the near future, several techniques such as 1) cryogenic distillation and selective absorption in solvents, 2) adsorption on activated charcoal, 3) selective sorption on chemical modified zeolites, or 4) diffusion through membranes with selective permeability are potential technologies to retain the gas.

Patricia Paviet-Hartmann; William Kerlin; Steven Bakhtiar

2010-11-01T23:59:59.000Z

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421

Synthesis, characterization, and ion exchange properties of a sodium nonatitanate, Na4Ti9O20.xH2O  

E-Print Network (OSTI)

During the Cold War, the Hanford Weapons Site in Richland, Washington, produced weapons grade plutonium which first needed to be separated from the other products using the PUREX process (plutonium and uranium extraction). As a by product of this process, millions of cubic meters of highly acidic radioactive waste were produced which are now stored in million gallon tanks at the Hanford site. Over the years, some tanks have been known to leak and some are even in danger of exploding. Because of these problems, the waste needs to be removed from these tanks and given permanent, safe storage. The purpose of this research is to produce a more efficient ion exchanger to separate the highly radioactive isotopes (9oSr, 137 Cs and transuranics) from the large quantities of inert salts. The smaller volume of high level waste produced can then be vitrified in glass and stored, while the low level waste can be poured into less expensive cement and glass. In this work, different parameters of the synthesis of the sodium nonatitanate ion exchanger, Na4Ti9O2OoxH20, such as the Na and Ti reactants, the heating time, oven temperature, Na:Ti mole ratio, and heating method, were altered and their effects on Sr2' ion exchange selectivity were examined. For example, the heating time was varied from I day to 2, 3, 7, and 30 days. Although the crystallinity remained the same from the I day to the 2 day sample, as the heating time further increased, the crystallinity improved. The most Sr selective material was the 2 day sample with a Kd (distribution coefficient) of 1.22x 106 MI/g in O.lM Na/ O.OOIM Sr solution. The Kd's steadily decreased as the sample crystallinity increased with a maximum Kd of only 1.6OxlO5 in O.OIM Na/ O.OO I M Sr solution after a heating time of 30 days. However, in a simulated waste such as NCAW, the 2 day sample gave a Kd of only 1.44x 105 MI/g, while the I day sample gave a value of 2.50x 105 . This indicates that the nonatitanate synthesis needs to be uniquely designed to optimize Sr 2+ removal in each specific type of waste to be remediated.

Graziano, Gina Marie

1998-01-01T23:59:59.000Z

422

Decontamination of high-level waste canisters  

SciTech Connect

This report presents evaluations of several methods for the in-process decontamination of metallic canisters containing any one of a number of solidified high-level waste (HLW) forms. The use of steam-water, steam, abrasive blasting, electropolishing, liquid honing, vibratory finishing and soaking have been tested or evaluated as potential techniques to decontaminate the outer surfaces of HLW canisters. Either these techniques have been tested or available literature has been examined to assess their applicability to the decontamination of HLW canisters. Electropolishing has been found to be the most thorough method to remove radionuclides and other foreign material that may be deposited on or in the outer surface of a canister during any of the HLW processes. Steam or steam-water spraying techniques may be adequate for some applications but fail to remove all contaminated forms that could be present in some of the HLW processes. Liquid honing and abrasive blasting remove contamination and foreign material very quickly and effectively from small areas and components although these blasting techniques tend to disperse the material removed from the cleaned surfaces. Vibratory finishing is very capable of removing the bulk of contamination and foreign matter from a variety of materials. However, special vibratory finishing equipment would have to be designed and adapted for a remote process. Soaking techniques take long periods of time and may not remove all of the smearable contamination. If soaking involves pickling baths that use corrosive agents, these agents may cause erosion of grain boundaries that results in rough surfaces.

Nesbitt, J.F.; Slate, S.C.; Fetrow, L.K.

1980-12-01T23:59:59.000Z

423

Idaho National Laboratory Description, Chellenges, Technology, Issues, and Needs  

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

i f th Hi i f th Hi h L l W t (HLW) P Overview of the High Level Waste (HLW) Program at the Id h N ti l L b t (INL) Sit Idaho National Laboratory (INL) Site Description Challenges Technology Issues and Needs Description, Challenges, Technology, Issues, and Needs 1 April 1, 2008 INL Site HLW is in Dry Storage in the Form of Calcine 8 9M ll f li id HLW t d t 4400 bi t f * 8-9M gallons of liquid HLW were converted to 4400 cubic meters of granular solid (calcine) through a fluidized bed calcination process - 7 to 1 volume reduction achieved * Average particle size is 0.4 cm * Bulk density is about 1 5 to 1 8 g/cc * Bulk density is about 1.5 to 1.8 g/cc - Contains roughly 44 metric tons heavy metal * Calcine is stored in 43 bins in 6 concrete-shielded binsets with one spare p - 7 th set of bins - intended for calcined SBW

424

Slide 1  

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

Inventory map reflects the non-federally owned SNF and HLW covered by the Nuclear Waste Policy Act Inventory map reflects the non-federally owned SNF and HLW covered by the Nuclear Waste Policy Act 2 Metric Tons Heavy Metal (MTHM) 3 Based on actual data through 2002 , as provided in the RW-859, and projected discharges for 2003-2010 which are rounded to two significant digits. Reflects trans-shipments as of end-2002. End of Year 2010 SNF & HLW Inventories 1 Approximately 64,000 MTHM 2 of Spent Nuclear Fuel (SNF) 3 & 275 High-Level Radioactive Waste (HLW) Canisters CT 1,900 TX 2,000 MD 1,200 VT 610 RI MT WY NE 790 SD ND OK KS 600 TX 2,000 LA 1,200 AR 1,200 IA 480 MN 1,100 WI 1,300 KY TN 1,500 MS 780 AL 3,000 GA 2,400 FL 2,900 NC 3,400 VA 2,400 WV OH 1,100 PA 5,800 ME 540 NJ 2,400 DE MI 2,500 MA 650 NH 480 IN SC 3,900 CO MO 670 IL 8,400 NY 3,300 CA 2,800 AZ 1,900 NM OR 360 NV UT WA 600 ID < 1 Commercial HLW 275 Canisters (~640 MTHM)

425

Regulatory Closure Options for the Residue in the Hanford Site Single-Shell Tanks  

SciTech Connect

Liquid, mixed, high-level radioactive waste (HLW) has been stored in 149 single-shell tanks (SSTS) located in tank farms on the U.S. Department of Energy's (DOE's) Hanford Site. The DOE is developing technologies to retrieve as much remaining HLW as technically possible prior to physically closing the tank farms. In support of the Hanford Tanks Initiative, Sandia National Laboratories has addressed the requirements for the regulatory closure of the radioactive component of any SST residue that may remain after physical closure. There is significant uncertainty about the end state of each of the 149 SSTS; that is, the nature and amount of wastes remaining in the SSTS after retrieval is uncertain. As a means of proceeding in the face of these uncertainties, this report links possible end-states with associated closure options. Requirements for disposal of HLW and low-level radioactive waste (LLW) are reviewed in detail. Incidental waste, which is radioactive waste produced incidental to the further processing of HLW, is then discussed. If the low activity waste (LAW) fraction from the further processing of HLW is determined to be incidental waste, then DOE can dispose of that incidental waste onsite without a license from the U.S. Nuclear Regulatory Commissions (NRC). The NRC has proposed three Incidental Waste Criteria for determining if a LAW fraction is incidental waste. One of the three Criteria is that the LAW fraction should not exceed the NRC's Class C limits.

Cochran, J.R. Shyr, L.J.

1998-10-05T23:59:59.000Z

426

Control Loop Tuning and Surge Response for Hanford WTP Melter Offgas Systems  

SciTech Connect

This report describes control loop tuning in models of the high level waste (HLW) melter offgas system, the low activity waste (LAW) melter offgas system and the HLW Pulse Jet Ventilation system and an assessment of the response to steam surges in both melter offgas systems. The three offgas systems were modeled using the Aspen Custom Modeler (ACM) software. The ACM models have been recently updated. Flowsheets of the system models used in this study are provided in Appendix D. To facilitate testing, these flowsheets represent somewhat simplified versions of the full models. For example, the HLW and LAW vessel ventilation systems have been represented as fixed air sources that provide a constant gas flow and specified air surges. Similarly, the six tanks and individual pulse-jet air sources in the HLW Pulse Jet Ventilation system are represented as a constant air source for control loop tuning purposes. The second LAW melter system has also been represented as a constant flow air source and several other simplifications such as removing HLW and LAW control interlocks, submerged bed scrubber bypass lines, and pressure relief valves have been made.

SMITH, FG III

2004-06-14T23:59:59.000Z

427

In-situ vitrification of soil. [Patent application  

DOE Patents (OSTI)

A method of vitrifying soil at or below a soil surface location. Two or more conductive electrodes are inserted into the soil for heating of the soil mass between them to a temperature above its melting temperature. Materials in the soil, such as buried waste, can thereby be effectively immobilized.

Brouns, R.A.; Buelt, J.L.; Bonner, W.F.

1981-04-06T23:59:59.000Z

428

Initial Laboratory-Scale Melter Test Results for Combined Fission Product Waste  

Science Conference Proceedings (OSTI)

This report describes the methods and results used to vitrify a baseline glass, CSLNTM-C-2.5 in support of the AFCI (Advanced Fuel Cycle Initiative) using a Quartz Crucible Scale Melter at the Pacific Northwest National Laboratory. Document number AFCI-WAST-PMO-MI-DV-2009-000184.

Riley, Brian J.; Crum, Jarrod V.; Buchmiller, William C.; Rieck, Bennett T.; Schweiger, Michael J.; Vienna, John D.

2009-10-01T23:59:59.000Z

429

Radioactive waste disposal package  

DOE Patents (OSTI)

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

Lampe, Robert F. (Bethel Park, PA)

1986-01-01T23:59:59.000Z

430

Method of making nanostructured glass-ceramic waste forms  

Science Conference Proceedings (OSTI)

A method of rendering hazardous materials less dangerous comprising trapping the hazardous material in nanopores of a nanoporous composite material, reacting the trapped hazardous material to render it less volatile/soluble, sealing the trapped hazardous material, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.

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

2012-12-18T23:59:59.000Z

431

High-Level Waste Corporate Board Performance Assessment Subcommittee  

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

Level Level Waste Corporate Board Performance Assessment Subcommittee John E. Marra, Ph.D. Associate Laboratory Director November 6, 2008 Richland, WA DOE-EM HLW Corporate Board Meeting Background - Performance Assessment Process Performance assessments are the fundamental risk assessment tool used by the DOE to evaluate and communicate the effectiveness and long-term impact of waste management and cleanup decisions. This includes demonstrations of compliance, NEPA analyses, and decisions about technologies and 2 analyses, and decisions about technologies and waste forms. Background - Process Perception EM-2 'Precepts' for Improved High-Level Waste Management (HLW Corporate Board Meeting - April 2008) Improved Performance Assessments (PA) The PA process is not consistently applied amongst the 3 The PA process is not consistently applied amongst the major HLW sites PA

432

 

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

Enhanced Chemical Cleaning (ECC) is a process for dissolving high-level waste (HLW) tank heel sludge with 1 wt% oxalic acid with pH controlled to 2 Enhanced Chemical Cleaning (ECC) is a process for dissolving high-level waste (HLW) tank heel sludge with 1 wt% oxalic acid with pH controlled to 2 followed by decomposition of the oxalic acid and separation of the precipitated solids by evaporation. The decomposition oxalic acid is accomplished in an external Advanced Oxidation Process (AOP) involving ozone injection and ultraviolet (UV) light exposure. Testing of the sludge dissolution, oxalate decomposition, precipitated solids separation, and downstream component solubility will be performed utilizing actual waste sludge from SRS HLW Tanks 5F, 12H, and 51H. Real-Waste Testing of Enhanced Chemical Cleaning for Sludge Heel Removal Savannah River Site Aiken South Carolina TC - A - 2010 - 020, Rev.0

433

ch_5  

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

45 45 DOE/EIS-0287 Idaho HLW & FD EIS 5.3.4.2 Existing Facilities Associated with High-Level Waste Management The facilities in this group are those that have historically been used at the INTEC to generate, treat, and store HLW. Because of the number of facilities involved, DOE has grouped them in functional groups for purposes of analysis (see Table 3-3). DOE analyzed the HLW tanks and bin sets for closure under all five disposition sce- narios; however, facilities that support the Tank Farm and bin sets were analyzed under a single disposition alternative. As shown in Table 3-3, the facility disposition alternative for most sup- porting facilities is Closure to Landfill Standards. (Two exceptions are the Liquid Effluent Treatment and Disposal Building and

434

 

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

SRNL will be receiving 4 replicate 1-Liter frit decon aqueous samples, and 2 replicate 1-Liter frit decon wet solids samples from DWPF. SRNL will be receiving 4 replicate 1-Liter frit decon aqueous samples, and 2 replicate 1-Liter frit decon wet solids samples from DWPF. These samples will be analyzed and the constituents compared to the Effluent Treatment Facility (ETF) Waste Acceptance Criteria (WAC). This work will take place in B-103 radiochem hood with sample submission to SRNL Analytical Department (AD) for analyses. These samples are all very low dose radioactive with whole body dose rates of nominal 0.01 mrem/hr. These samples are generated from frit blasting of the outside of the DWPF High Level Waste (HLW) steel canisters that contain HLW glass. The frit blasting generates residual frit that is washed/slurried with water during the outer HLW container frit blasting process.

435

Page not found | Department of Energy  

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

51 - 11860 of 26,764 results. 51 - 11860 of 26,764 results. Article U.S. Energy Secretary Chu to Lead Delegation to IAEA 54th Annual General Conference http://energy.gov/articles/us-energy-secretary-chu-lead-delegation-iaea-54th-annual-general-conference Download Smart Grid Conceptual Actors/Data Flow Diagram- Cross Domain Network Focued- Open SG/SG-Network TF http://energy.gov/gc/downloads/smart-grid-conceptual-actorsdata-flow-diagram-cross-domain-network-focued-open-sgsg Download End of Year 2010 SNF & HLW Inventories Map of the United States of America that shows the location of approximately 64,000 MTHM of Spent Nuclear Fuel (SNF) & 275 High-Level Radioactive Waste (HLW) Canisters. http://energy.gov/gc/downloads/end-year-2010-snf-hlw-inventories Download Microsoft PowerPoint- PARS II CPP Deployment Schedule 13Aug10.pptx

436

Tank Waste Corporate Board Meeting 11/06/08 | Department of Energy  

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

11/06/08 11/06/08 Tank Waste Corporate Board Meeting 11/06/08 The following documents are associated with the Tank Waste Corporate Board Meeting held on November 6th, 2008. Note: (Please contact Steven Ross at steven.ross@em.doe.gov for a HLW Glass Waste Loadings version with animations on slide 6). Slurry Retrieval, Pipeline Transport & Plugging and Mixing Workshop The Way Ahead - West Valley Demonstration Project High-Level Liquid Waste Tank Integrity Workshop - 2008 Savannah River Tank Waste Residuals Hanford Tank Waste Residuals HLW Glass Waste Loadings High-Level Waste Corporate Board Performance Assessment Subcommittee More Documents & Publications Tank Waste Corporate Board Meeting 11/18/10 System Planning for Low-Activity Waste at Hanford Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility

437

 

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

Enhanced Chemical Cleaning (ECC) is a process for dissolving high-level waste (HLW) tank heel sludge with 2 wt% oxalic acid followed by Enhanced Chemical Cleaning (ECC) is a process for dissolving high-level waste (HLW) tank heel sludge with 2 wt% oxalic acid followed by decomposition of the oxalic acid and separation of the precipitated solids by evaporation. The decomposition oxalic acid is accomplished in an external Advanced Oxidation Process (AOP) involving ozone injection both with and without ultraviolet (UV) light exposure. Testing of the sludge dissolution, oxalate decomposition, downstream component solubility, and solids rheology, settling rate, and particle size will be perform