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1

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

SciTech Connect (OSTI)

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

Mayberry, J.L.; DeWitt, L.M. [Science Applications International Corp., Idaho Falls, ID (United States); Darnell, R. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

1993-08-01T23:59:59.000Z

2

Electron Microscopy Characterization of Tc-Bearing Metallic Waste Forms- Final Report FY10  

SciTech Connect (OSTI)

The DOE Fuel Cycle Research & Development (FCR&D) Program is developing aqueous and electrochemical approaches to the processing of used nuclear fuel that will generate technetium-bearing waste streams. This final report presents Pacific Northwest National Laboratory (PNNL) research in FY10 to evaluate an iron-based alloy waste form for Tc that provides high waste loading within waste form processing limitations, meets waste form performance requirements for durability and the long-term retention of radionuclides and can be produced with consistent physical, chemical, and radiological properties that meet regulatory acceptance requirements for disposal.

Buck, Edgar C.; Neiner, Doinita

2010-09-30T23:59:59.000Z

3

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

SciTech Connect (OSTI)

This document defines the product performance criteria to be used in Phase I of the Final Waste Forms Project. In Phase I, treatability studies will be performed to provide {open_quotes}proof-of-principle{close_quotes} data to establish the viability of stabilization/solidification (S/S) technologies. This information is required by March 1995. In Phase II, further treatability studies, some at the pilot scale, will be performed to provide sufficient data to allow treatment alternatives identified in Phase I to be more fully developed and evaluated, as well as to reduce performance uncertainties for those methods chosen to treat a specific waste. Three main factors influence the development and selection of an optimum waste form formulation and hence affect selection of performance criteria. These factors are regulatory, process-specific, and site-specific waste form standards or requirements. Clearly, the optimum waste form formulation will require consideration of performance criteria constraints from each of the three categories. Phase I will focus only on the regulatory criteria. These criteria may be considered the minimum criteria for an acceptable waste form. In other words, a S/S technology is considered viable only if it meet applicable regulatory criteria. The criteria to be utilized in the Phase I treatability studies were primarily taken from Environmental Protection Agency regulations addressed in 40 CFR 260 through 265 and 268; and Nuclear Regulatory Commission regulations addressed in 10 CFR 61. Thus the majority of the identified criteria are independent of waste form matrix composition (i.e., applicable to cement, glass, organic binders etc.).

Gilliam, T.M. [Oak Ridge National Lab., TN (United States); Hutchins, D.A. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States); Chodak, P. III [Massachusetts Institute of Technology (United States)

1994-06-01T23:59:59.000Z

4

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

SciTech Connect (OSTI)

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

5

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

SciTech Connect (OSTI)

The end of the Cold War raised the need for the technical community to be concerned with the disposition of excess nuclear weapon material. The plutonium will either be converted into mixed-oxide fuel for use in nuclear reactors or immobilized in glass or ceramic waste forms and placed in a repository. The stability and behavior of plutonium in the ceramic materials as well as the phase behavior and stability of the ceramic material in the environment is not well established. In order to provide technically sound solutions to these issues, thermodynamic data are essential in developing an understanding of the chemistry and phase equilibria of the actinide-bearing mineral waste form materials proposed as immobilization matrices. Mineral materials of interest include zircon, zirconolite, and pyrochlore. High temperature solution calorimetry is one of the most powerful techniques, sometimes the only technique, for providing the fundamental thermodynamic data needed to establish optimum material fabrication parameters, and more importantly understand and predict the behavior of the mineral materials in the environment. The purpose of this project is to experimentally determine the enthalpy of formation of actinide orthosilicates, the enthalpies of formation of actinide substituted zirconolite and pyrochlore, and develop an understanding of the bonding characteristics and stabilities of these materials.

Williamson, Mark A.; Ebbinghaus, Bartley B.; Navrotsky, Alexandra

2001-03-01T23:59:59.000Z

6

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

SciTech Connect (OSTI)

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

7

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

SciTech Connect (OSTI)

This report presents a preliminary parametric performance assessment (PA) of potential waste disposal systems for alpha-contaminated, mixed, low-level waste (ALLW) currently stored at the Transuranic Storage Area of INEL. The ALLW, which contains from 10 to 100 nCi/g of transuranic (TRU) radionuclides, is awaiting treatment and disposal. The purpose of this study was to examine the effects of several parameters on the radiological-confinement performance of potential disposal systems for the ALLW. The principal emphasis was on the performance of final waste forms (FWFs). Three categories of FWF (cement, glass, and ceramic) were addressed by evaluating the performance of two limiting FWFs for each category. Performance at five conceptual disposal sites was evaluated to illustrate the effects of site characteristics on the performance of the total disposal system. Other parameters investigated for effects on receptor dose included inventory assumptions, TRU radionuclide concentration, FWF fracture, disposal depth, water infiltration rates, subsurface-transport modeling assumptions, receptor well location, intrusion scenario assumptions, and the absence of waste immobilization. These and other factors were varied singly and in some combinations. The results indicate that compliance of the treated and disposed ALLW with the performance objectives depends on the assumptions made, as well as on the FWF and the disposal site. Some combinations result in compliance, while others do not. The implications of these results for decision making relative to treatment and disposal of the INEL ALLW are discussed. The report compares the degree of conservatism in this preliminary parametric PA against that in four other PAs and one risk assessment. All of the assessments addressed the same disposal site, but different wastes. The report also presents a qualitative evaluation of the uncertainties in the PA and makes recommendations for further study.

Smith, T.H.; Sussman, M.E. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); Myers, J.; Djordjevic, S.M.; DeBiase, T.A.; Goodrich, M.T.; DeWitt, D. [IT Corp., Albuquerque, NM (United States)

1995-08-01T23:59:59.000Z

8

Investigation of microscopic radiation damage in waste forms using ODNMR and AEM techniques. (EMSP Project Final Report)  

SciTech Connect (OSTI)

This project seeks to understand the microscopic effects of radiation damage in nuclear waste forms. The authors' approach to this challenge encompasses studies of ceramics and glasses containing short-lived alpha- and beta-emitting actinides with electron microscopy, laser and X-ray spectroscopic techniques, and computational modeling and simulations. In order to obtain information on long-term radiation effects on waste forms, much of the effort is to investigate {alpha}-decay induced microscopic damage in 18-year old samples of crystalline yttrium and lutetium orthophosphates that initially contained {approximately} 1(wt)% of the alpha-emitting isotope {sup 244}Cm (18.1 y half life). Studies also are conducted on borosilicate glasses that contain {sup 244}Cm, {sup 241}Am, or {sup 249}Bk, respectively. The authors attempt to gain clear insights into the properties of radiation-induced structure defects and the consequences of collective defect-environment interactions, which are critical factors in assessing the long-term performance of high-level nuclear waste forms.

Liu, G.; Luo, J.; Beitz, J.; Li, S.; Williams, C.; Zhorin, V.

2000-04-21T23:59:59.000Z

9

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

SciTech Connect (OSTI)

The U.S. Department of Energy’s Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford’s tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as {sup 137}Cs, {sup 129}I, {sup 99}Tc, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150°C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750°C) continuous method by which WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage, but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the SRNL to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. BSR testing with WTP SW waste surrogates and associated analytical analyses and tests of granular products (GP) and monoliths began in the Fall of 2009, and then was continued from the Fall of 2010 through the Spring of 2011. Radioactive testing commenced in 2010 with a demonstration of Hanford’s WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of {sup 125/129}I and {sup 99}Tc to chemically resemble WTP-SW. Prior to these radioactive feed tests, non-radioactive simulants were also processed. Ninety six grams of radioactive granular product were made for testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing. The granular products (both simulant and radioactive) were tested and a subset of the granular material (both simulant and radioactive) were stabilized in a geopolymer matrix. Extensive testing and characterization of the granular and monolith material were made including the following: ? ASTM C1285 (Product Consistency Test) testing of granular and monolith; ? ASTM C1308 accelerated leach testing of the radioactive monolith; ? ASTM C192 compression testing of monoliths; and ? EPA Method 1311 Toxicity Characteristic Leaching Procedure (TCLP) testing. The significant findings of the testing completed on simulant and radioactive WTP-SW are given below: ? Data indicates {sup 99}Tc, Re, Cs, and I

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

2014-08-21T23:59:59.000Z

10

Process for immobilizing plutonium into vitreous ceramic waste forms  

DOE Patents [OSTI]

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

Feng, X.; Einziger, R.E.

1997-08-12T23:59:59.000Z

11

Process for immobilizing plutonium into vitreous ceramic waste forms  

DOE Patents [OSTI]

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

Feng, X.; Einziger, R.E.

1997-01-28T23:59:59.000Z

12

The Measurement of Thermal Diffusivity of Simulated Glass Forming Nuclear Waste Melts  

Science Journals Connector (OSTI)

High-level nuclear waste is generated during reprocessing of nuclear reactor fuels. At present, these wastes are stored at various locations in the United States until a final waste form (i.e., glass, SYNROC, ......

James U. Derby; L. David Pye; M. J. Plodinec

1983-01-01T23:59:59.000Z

13

Retrieval Of Final Stored Radioactive Waste Resumes  

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

18, 2012 18, 2012 Media Contact: Danielle Miller, DOE-Idaho Operations, 208-526-5709, millerdc@id.doe.gov Rick Dale, Idaho Treatment Group, 208-557-6552, rick.dale@amwtp.inl.gov Retrieval Of Final Stored Radioactive Waste Resumes IDAHO FALLS, ID- Operations to retrieve the estimated 6,900 cubic meters of stored transuranic waste remaining at the Idaho site began this week at the U.S. Department of Energy�s Advanced Mixed Waste Treatment Project. Waste retrieval resumes at the Advanced Mixed Waste Treatment Project. The resumption of work comes after a nearly two-year stoppage of retrieval operations �A significant investment has been made in terms of time and dollars that will allow employees to safely retrieve the final radioactive waste that has been stored aboveground at the Idaho site for more than four

14

CERAMIC WASTE FORM DATA PACKAGE  

SciTech Connect (OSTI)

The purpose of this data package is to provide information about simulated crystalline waste forms that can be used to select an appropriate composition for a Cold Crucible Induction Melter (CCIM) proof of principle demonstration. Melt processing, viscosity, electrical conductivity, and thermal analysis information was collected to assess the ability of two potential candidate ceramic compositions to be processed in the Idaho National Laboratory (INL) CCIM and to guide processing parameters for the CCIM operation. Given uncertainties in the CCIM capabilities to reach certain temperatures throughout the system, one waste form designated 'Fe-MP' was designed towards enabling processing and another, designated 'CAF-5%TM-MP' was designed towards optimized microstructure. Melt processing studies confirmed both compositions could be poured from a crucible at 1600{degrees}C although the CAF-5%TM-MP composition froze before pouring was complete due to rapid crystallization (upon cooling). X-ray diffraction measurements confirmed the crystalline nature and phase assemblages of the compositions. The kinetics of melting and crystallization appeared to vary significantly between the compositions. Impedance spectroscopy results indicated the electrical conductivity is acceptable with respect to processing in the CCIM. The success of processing either ceramic composition will depend on the thermal profiles throughout the CCIM. In particular, the working temperature of the pour spout relative to the bulk melter which can approach 1700{degrees}C. The Fe-MP composition is recommended to demonstrate proof of principle for crystalline simulated waste forms considering the current configuration of INL's CCIM. If proposed modifications to the CCIM can maintain a nominal temperature of 1600{degrees}C throughout the melter, drain, and pour spout, then the CAF-5%TM-MP composition should be considered for a proof of principle demonstration.

Amoroso, J.; Marra, J.

2014-06-13T23:59:59.000Z

15

Mixed Waste Salt Encapsulation Using Polysiloxane - Final Report  

SciTech Connect (OSTI)

A proof-of-concept experimental study was performed to investigate the use of Orbit Technologies polysiloxane grouting material for encapsulation of U.S. Department of Energy mixed waste salts leading to a final waste form for disposal. Evaporator pond salt residues and other salt-like material contaminated with both radioactive isotopes and hazardous components are ubiquitous in the DOE complex and may exceed 250,000,000 kg of material. Current treatment involves mixing low waste percentages (less than 10% by mass salt) with cement or costly thermal treatment followed by cementation to the ash residue. The proposed technology involves simple mixing of the granular salt material (with relatively high waste loadings-greater than 50%) in a polysiloxane-based system that polymerizes to form a silicon-based polymer material. This study involved a mixing study to determine optimum waste loadings and compressive strengths of the resultant monoliths. Following the mixing study, durability testing was performed on promising waste forms. Leaching studies including the accelerated leach test and the toxicity characteristic leaching procedure were also performed on a high nitrate salt waste form. In addition to this testing, the waste form was examined by scanning electron microscope. Preliminary cost estimates for applying this technology to the DOE complex mixed waste salt problem is also given.

Miller, C.M.; Loomis, G.G.; Prewett, S.W.

1997-11-01T23:59:59.000Z

16

Miscellaneous Waste-Form FEPs  

SciTech Connect (OSTI)

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

A. Schenker

2000-12-08T23:59:59.000Z

17

Qualifying radioactive waste forms for geologic disposal  

SciTech Connect (OSTI)

We have developed a phased strategy that defines specific program-management activities and critical documentation for producing radioactive waste forms, from pyrochemical processing of spent nuclear fuel, that will be acceptable for geologic disposal by the US Department of Energy. The documentation of these waste forms begins with the decision to develop the pyroprocessing technology for spent fuel conditioning and ends with production of the last waste form for disposal. The need for this strategy is underscored by the fact that existing written guidance for establishing the acceptability for disposal of radioactive waste is largely limited to borosilicate glass forms generated from the treatment of aqueous reprocessing wastes. The existing guidance documents do not provide specific requirements and criteria for nonstandard waste forms such as those generated from pyrochemical processing operations.

Jardine, L.J. [Lawrence Livermore National Lab., CA (United States); Laidler, J.J.; McPheeters, C.C. [Argonne National Lab., IL (United States)

1994-09-01T23:59:59.000Z

18

Waste reduction through consumer education. Final report  

SciTech Connect (OSTI)

The Waste Reduction through Consumer Education research project was conducted to determine how environmental educational strategies influence purchasing behavior in the supermarket. The objectives were to develop, demonstrate, and evaluate consumer education strategies for waste reduction. The amount of waste generated by packaging size and form, with an adjustment for local recyclability of waste, was determined for 14 product categories identified as having more waste generating and less waste generating product choices (a total of 484 products). Using supermarket scan data and shopper identification numbers, the research tracked the purchases of shoppers in groups receiving different education treatments for 9 months. Statistical tests applied to the purchase data assessed patterns of change between the groups by treatment period. Analysis of the data revealed few meaningful statistical differences between study groups or changes in behavior over time. Findings suggest that broad brush consumer education about waste reduction is not effective in changing purchasing behaviors in the short term. However, it may help create a general awareness of the issues surrounding excess packaging and consumer responsibility. The study concludes that the answer to waste reduction in the future may be a combination of voluntary initiatives by manufacturers and retailers, governmental intervention, and better-informed consumers.

Harrison, E.Z.

1996-05-01T23:59:59.000Z

19

DWPF waste form compliance plan (Draft Revision)  

SciTech Connect (OSTI)

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

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

1991-12-31T23:59:59.000Z

20

DWPF waste form compliance plan (Draft Revision)  

SciTech Connect (OSTI)

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

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

1991-01-01T23:59:59.000Z

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21

EIS-0287: Idaho High-Level Waste and Facilities Disposition Final  

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

Idaho High-Level Waste and Facilities Disposition Final Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002) EIS-0287: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002) This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liquid waste at the Idaho National Engineering and Environmental Laboratory (INEEL) in liquid and solid forms. This EIS also analyzes alternatives for the final disposition of HLW management facilities at the INEEL after their missions are completed. Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, DOE/EIS-0287 (September 2002)

22

Phosphates as Nuclear Waste Forms  

Science Journals Connector (OSTI)

...environment of the disposal site, the...the sustained funding of the Office...EP (1999) Yucca Mountain as a radioactive-waste...Ultimate disposal of radioactive...Adirondack Mountains, New York...for geologic disposal. Mater Res...

Rodney C. Ewing; LuMin Wang

23

Retrieval Of Final Stored Radioactive Waste Resumes | Department...  

Office of Environmental Management (EM)

Retrieval Of Final Stored Radioactive Waste Resumes Retrieval Of Final Stored Radioactive Waste Resumes April 18, 2012 - 12:00pm Addthis Media Contacts Danielle Miller, DOE-Idaho...

24

Mixed low-level waste form evaluation  

SciTech Connect (OSTI)

A scoping level evaluation of polyethylene encapsulation and vitreous waste forms for safe storage of mixed low-level waste was performed. Maximum permissible radionuclide concentrations were estimated for 15 indicator radionuclides disposed of at the Hanford and Savannah River sites with respect to protection of the groundwater and inadvertent intruder pathways. Nominal performance improvements of polyethylene and glass waste forms relative to grout are reported. These improvements in maximum permissible radionuclide concentrations depend strongly on the radionuclide of concern and pathway. Recommendations for future research include improving the current understanding of the performance of polymer waste forms, particularly macroencapsulation. To provide context to these estimates, the concentrations of radionuclides in treated DOE waste should be compared with the results of this study to determine required performance.

Pohl, P.I.; Cheng, Wu-Ching; Wheeler, T.; Waters, R.D.

1997-03-01T23:59:59.000Z

25

DUF6 Final EIS Document Request Form  

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

EIS Request Form EIS Request Form Final EIS Document Request Form Use the form below to order copies of the DUF6 Conversion Facility Final EISs and Records of Decision. Step 3 Select the EIS that you want to receive. Select one of the three options below. My request applies to the Paducah Conversion Facility EIS My request applies to the Portsmouth Conversion Facility EIS My request applies to both the Paducah and the Portsmouth Conversion Facility EISs Step 1 Request EIS copies. Choose one or more of the following: Mail me a compact disc (CD-ROM) of the Final EIS and Record of Decision. Mail me a printed copy of the Final EIS and Record of Decision. Step 2 Enter your personal information. You must submit your full name and complete address including zip code to receive postal mail. You must provide an email address if you want to receive email notifications.

26

Secondary Waste Forms and Technetium Management  

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

Secondary Waste Forms and Secondary Waste Forms and Technetium Management Joseph H. Westsik, Jr. Pacific Northwest National Laboratory EM HLW Corporate Board Meeting November 18, 2010 What are Secondary Wastes? Process condensates and scrubber and/or off-gas treatment liquids from the pretreatment and ILAW melter facilities at the Hanford WTP. Sent from WTP to the Effluent Treatment Facility (ETF) for treatment and disposal Treated liquid effluents under the ETF State Wastewater Discharge Permit Solidified liquid effluents under the Dangerous Waste Permit for disposal at the Integrated Disposal Facility (IDF) Solidification Treatment Unit to be added to ETF to provide capacity for WTP secondary liquid wastes 2 Evaporator Condensate Solution Evaporator Pretreatment Melter SBS/ WESP Secondary

27

Alternative Waste Forms for Electro-Chemical Salt Waste  

SciTech Connect (OSTI)

This study was undertaken to examine alternate crystalline (ceramic/mineral) and glass waste forms for immobilizing spent salt from the Advanced Fuel Cycle Initiative (AFCI) electrochemical separations process. The AFCI is a program sponsored by U.S. Department of Energy (DOE) to develop and demonstrate a process for recycling spent nuclear fuel (SNF). The electrochemical process is a molten salt process for the reprocessing of spent nuclear fuel in an electrorefiner and generates spent salt that is contaminated with alkali, alkaline earths, and lanthanide fission products (FP) that must either be cleaned of fission products or eventually replaced with new salt to maintain separations efficiency. Currently, these spent salts are mixed with zeolite to form sodalite in a glass-bonded waste form. The focus of this study was to investigate alternate waste forms to immobilize spent salt. On a mole basis, the spent salt is dominated by alkali and Cl with minor amounts of alkaline earth and lanthanides. In the study reported here, we made an effort to explore glass systems that are more compatible with Cl and have not been previously considered for use as waste forms. In addition, alternate methods were explored with the hope of finding a way to produce a sodalite that is more accepting of as many FP present in the spent salt as possible. This study was done to investigate two different options: (1) alternate glass families that incorporate increased concentrations of Cl; and (2) alternate methods to produce a mineral waste form.

Crum, Jarrod V.; Sundaram, S. K.; Riley, Brian J.; Matyas, Josef; Arreguin, Shelly A.; Vienna, John D.

2009-10-28T23:59:59.000Z

28

Field testing of waste forms using lysimeters  

SciTech Connect (OSTI)

The Low-Level Waste Data Base Development - EPICOR-II Resin/Liner Investigation Program funded by the US Nuclear Regulatory Commission is obtaining information on performance of radioactive waste in a disposal environment. Waste forms manufactured from ion exchange resins used to clean up water from the accident at Three Mile Island Nuclear Power Station are being examined in field tests. This paper presents a description of the field testing and results from the first year of operation. 8 refs., 8 figs., 4 tabs.

McConnell, J.W. Jr.; Rogers, R.D.

1987-01-01T23:59:59.000Z

29

The Ceramic Waste Form Process at Idaho National Laboratory  

SciTech Connect (OSTI)

The treatment of spent nuclear fuel for disposition using an electrometallurgical technique results in two high-level waste forms: a ceramic waste form (CWF) and a metal waste form. Reactive metal fuel constituents, including all the transuranic metals and the majority of the fission products remain in the salt as chlorides and are processed into the CWF. The solidified salt is containerized and transferred to the CWF process where it is ground in an argon atmosphere. Zeolite 4A is ground and then dried in a mechanically-fluidized dryer. The salt and zeolite are mixed in a V-mixer and heated to 500°C to occlude the salt into the structure of the zeolite. The salt-loaded zeolite is cooled, mixed with borosilicate glass frit, and transferred to a crucible, which is placed in a furnace and heated to 925°C. During this process, known as pressureless consolidation, the zeolite is converted to the final sodalite form and the glass thoroughly encapsulates the sodalite, producing a dense, leach-resistant final waste form.

Stephen Priebe

2007-05-01T23:59:59.000Z

30

NNSA CX Determination Form (Final)(3).pdf  

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

Nuclear Security Administration Nuclear Security Administration Categorical Exclusion Determination Form _____________________________________________________________________________________________ NEPA ID #: Project Title: Program or Field Office: Location: _____________________________________________________________________________________________ Proposed Action or Project Description: _____________________________________________________________________________________________ This action would not: threaten a violation of applicable statutory, regulatory, or permit requirements for environment, safety, and health, including DOE and/or Executive Orders; require siting, construction, or major expansion of waste storage, disposal, recovery, or treatment facilities, but may include such categorically

31

Engineering-Scale Demonstration of DuraLith and Ceramicrete Waste Forms  

SciTech Connect (OSTI)

To support the selection of a waste form for the liquid secondary wastes from the Hanford Waste Immobilization and Treatment Plant, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing on four candidate waste forms. Two of the candidate waste forms have not been developed to scale as the more mature waste forms. This work describes engineering-scale demonstrations conducted on Ceramicrete and DuraLith candidate waste forms. Both candidate waste forms were successfully demonstrated at an engineering scale. A preliminary conceptual design could be prepared for full-scale production of the candidate waste forms. However, both waste forms are still too immature to support a detailed design. Formulations for each candidate waste form need to be developed so that the material has a longer working time after mixing the liquid and solid constituents together. Formulations optimized based on previous lab studies did not have sufficient working time to support large-scale testing. The engineering-scale testing was successfully completed using modified formulations. Further lab development and parametric studies are needed to optimize formulations with adequate working time and assess the effects of changes in raw materials and process parameters on the final product performance. Studies on effects of mixing intensity on the initial set time of the waste forms are also needed.

Josephson, Gary B.; Westsik, Joseph H.; Pires, Richard P.; Bickford, Jody; Foote, Martin W.

2011-09-23T23:59:59.000Z

32

Proposed research and development plan for mixed low-level waste forms  

SciTech Connect (OSTI)

The objective of this report is to recommend a waste form program plan that addresses waste form issues for mixed low-level waste (MLLW). The report compares the suitability of proposed waste forms for immobilizing MLLW in preparation for permanent near-surface disposal and relates them to their impact on the U.S. Department of Energy`s mixed waste mission. Waste forms are classified into four categories: high-temperature waste forms, hydraulic cements, encapsulants, and specialty waste forms. Waste forms are evaluated concerning their ability to immobilize MLLW under certain test conditions established by regulatory agencies and research institutions. The tests focused mainly on leach rate and compressive strength. Results indicate that all of the waste forms considered can be tailored to give satisfactory performance immobilizing large fractions of the Department`s MLLW inventory. Final waste form selection will ultimately be determined by the interaction of other, often nontechnical factors, such as economics and politics. As a result of this report, three top-level programmatic needs have been identified: (1) a basic set of requirements for waste package performance and disposal; (2) standardized tests for determining waste form performance and suitability for disposal; and (3) engineering experience operating production-scale treatment and disposal systems for MLLW.

O`Holleran, T.O.; Feng, X.; Kalb, P. [and others

1996-12-01T23:59:59.000Z

33

Final Hanford Offsite Waste Shipment Leaves Idaho Treatment Facility |  

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

Final Hanford Offsite Waste Shipment Leaves Idaho Treatment Final Hanford Offsite Waste Shipment Leaves Idaho Treatment Facility Final Hanford Offsite Waste Shipment Leaves Idaho Treatment Facility August 18, 2011 - 12:00pm Addthis Idaho State Patrol Troopers Rick Stouse and Tony Anderson inspected the TRUPACTS, containers which contain TRU waste, and trailer containing the final shipment of Hanford offsite waste. The Idaho State Patrol officers have played an important role in AMWTP's success by inspecting every one of AMWTP's nearly 3,900 shipments. Idaho State Patrol Troopers Rick Stouse and Tony Anderson inspected the TRUPACTS, containers which contain TRU waste, and trailer containing the final shipment of Hanford offsite waste. The Idaho State Patrol officers have played an important role in AMWTP's success by inspecting every one of

34

Microbial degradation of low-level radioactive waste. Final report  

SciTech Connect (OSTI)

The Nuclear Regulatory Commission stipulates in 10 CFR 61 that disposed low-level radioactive waste (LLW) be stabilized. To provide guidance to disposal vendors and nuclear station waste generators for implementing those requirements, the NRC developed the Technical Position on Waste Form, Revision 1. That document details a specified set of recommended testing procedures and criteria, including several tests for determining the biodegradation properties of waste forms. Information has been presented by a number of researchers, which indicated that those tests may be inappropriate for examining microbial degradation of cement-solidified LLW. Cement has been widely used to solidify LLW; however, the resulting waste forms are sometimes susceptible to failure due to the actions of waste constituents, stress, and environment. The purpose of this research program was to develop modified microbial degradation test procedures that would be more appropriate than the existing procedures for evaluation of the effects of microbiologically influenced chemical attack on cement-solidified LLW. The procedures that have been developed in this work are presented and discussed. Groups of microorganisms indigenous to LLW disposal sites were employed that can metabolically convert organic and inorganic substrates into organic and mineral acids. Such acids aggressively react with cement and can ultimately lead to structural failure. Results on the application of mechanisms inherent in microbially influenced degradation of cement-based material are the focus of this final report. Data-validated evidence of the potential for microbially influenced deterioration of cement-solidified LLW and subsequent release of radionuclides developed during this study are presented.

Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W. Jr

1996-06-01T23:59:59.000Z

35

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

Open Energy Info (EERE)

application (Form 200) Jump to: navigation, search OpenEI Reference LibraryAdd to library Form: Report of Waste Discharge application (Form 200) Abstract Persons discharging or...

36

Revision 08 (08/10) Form G Radioactive Waste Disposal Form  

E-Print Network [OSTI]

Revision 08 (08/10) Form G Radioactive Waste Disposal Form RS - 19g Proc. 9290, 9501 General Instructions: 1. Do not mix different waste forms together. Keep dry, liquid, and scintillation vials separate. 2. Do not mix waste of different isotopes. 3. Entries are to be made on this form each time waste

Nair, Sankar

37

Final Transuranic Waste Shipment Leaves Rocky Flats | Department of Energy  

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

Final Transuranic Waste Shipment Leaves Rocky Flats Final Transuranic Waste Shipment Leaves Rocky Flats Final Transuranic Waste Shipment Leaves Rocky Flats April 19, 2005 - 12:23pm Addthis Cleanup Ahead of Schedule, On Track to Save Taxpayers Billions GOLDEN, CO. - A major environmental victory was achieved at the Rocky Flats Site in Golden, Colo., today when the final remaining shipment of radioactive, transuranic (TRU) waste left the property on a truck bound for an underground waste repository in New Mexico. This major milestone is another step toward the final conversion of the site to a National Wildlife Refuge managed by the U.S. Fish and Wildlife Service. "This is great news for all of Colorado, and would not have been possible without hand-in-glove cooperation between the Department of Energy, the

38

Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview  

SciTech Connect (OSTI)

In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified.

Dayal, R.; Lee, B.S.; Wilke, R.J.; Swyler, K.J.; Soo, P.; Ahn, T.M.; McIntyre, N.S.; Veakis, E.

1982-02-01T23:59:59.000Z

39

Issuance of the Final Tank Closure and Waste Management Environmental  

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

Issuance of the Final Tank Closure and Waste Management Issuance of the Final Tank Closure and Waste Management Environmental Impact Statement Issuance of the Final Tank Closure and Waste Management Environmental Impact Statement December 5, 2012 - 12:00pm Addthis Media Contacts Carrie Meyer, DOE (509) 376-0810 Carrie_C_Meyer@orp.doe.gov Erika Holmes, Ecology (509) 372-7880 Erika.Holmes@ecy.wa.gov Richland, WA - The U.S. Department of Energy (DOE) is issuing its Final Tank Closure and Waste Management Environmental Impact Statement Hanford Site, Richland, Washington" (Final TC & WM EIS, DOE/EIS-0391), prepared in accordance with the National Environmental Policy Act (NEPA). The Environmental Protection Agency (EPA) and Washington State Department of Ecology are cooperating agencies on this Final EIS, which analyzes

40

RCRA Notification of Regulated Waste Activity (EPA Form 8700...  

Open Energy Info (EERE)

Notification of Regulated Waste Activity (EPA Form 8700-12) Jump to: navigation, search OpenEI Reference LibraryAdd to library Form: RCRA Notification of Regulated Waste Activity...

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


41

The Ceramic Waste Form Process at the Idaho National Laboratory  

SciTech Connect (OSTI)

The treatment of spent nuclear fuel for disposition using an electrometallurgical technique results in two high-level waste forms: a ceramic waste form (CWF) and a metal waste form (MWF). The CWF is a composite of sodalite and glass, which stabilizes the active fission products (alkali, alkaline earths, and rare earths) and transuranic (TRU) elements. Reactive metal fuel constituents, including all the TRU metals and the majority of the fission products remain in the salt as chlorides and are processed into the CWF. The solidified salt is containerized and transferred to the CWF process where it is ground in an argon atmosphere. Zeolite 4A is dried in a mechanically-fluidized dryer to about 0.1 wt% moisture and ground to a particle-size range of 45µ to 250µ. The salt and zeolite are mixed in a V-mixer and heated to 500°C for about 18 hours. During this process, the salt occludes into the structure of the zeolite. The salt-loaded zeolite (SLZ) is cooled and then mixed with borosilicate glass frit with a comparable particle-size range. The SLZ/glass mixture is transferred to a crucible, which is placed in a furnace and heated to 925°C. During this process, known as pressureless consolidation, the zeolite is converted to the final sodalite form and the glass thoroughly encapsulates the sodalite, producing a dense, leach-resistant final waste form. During the last several years, changes have occurred to the process, including: particle size of input materials and conversion from hot isostatic pressing to pressureless consolidation, This paper is intended to provide the current status of the CWF process focusing on the adaptation to pressureless consolidation. Discussions will include impacts of particle size on final waste form and the pressureless consolidation cycle. A model will be presented that shows the heating and cooling cycles and the effect of radioactive decay heat on the amount of fission products that can be incorporated into the CWF.

Ken Bateman; Stephen Priebe

2006-08-01T23:59:59.000Z

42

DURABILITY TESTING OF FLUIDIZED BED STEAM REFORMER (FBSR) WASTE FORMS  

SciTech Connect (OSTI)

Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium aqueous radioactive wastes. The addition of clay and a catalyst as co-reactants converts high sodium aqueous low activity wastes (LAW) such as those existing at the Hanford and Idaho DOE sites to a granular ''mineralized'' waste form that may be made into a monolith form if necessary. Simulant Hanford and Idaho high sodium wastes were processed in a pilot scale FBSR at Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW). The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The durability of the FBSR waste form products was tested in order to compare the measured durability to previous FBSR waste form testing on Hanford Envelope C waste forms that were made by THOR Treatment Technologies (TTT) and to compare the FBSR durability to vitreous LAW waste forms, specifically the Hanford low activity waste (LAW) glass known as the Low-activity Reference Material (LRM). The durability of the FBSR waste form is comparable to that of the LRM glass for the test responses studied.

Jantzen, C

2006-01-06T23:59:59.000Z

43

Electrochemical corrosion testing of metal waste forms  

SciTech Connect (OSTI)

Electrochemical corrosion tests have been conducted on simulated stainless steel-zirconium (SS-Zr) metal waste form (MWF) samples. The uniform aqueous corrosion behavior of the samples in various test solutions was measured by the polarization resistance technique. The data show that the MWF corrosion rates are very low in groundwaters representative of the proposed Yucca Mountain repository. Galvanic corrosion measurements were also conducted on MWF samples that were coupled to an alloy that has been proposed for the inner lining of the high-level nuclear waste container. The experiments show that the steady-state galvanic corrosion currents are small. Galvanic corrosion will, hence, not be an important mechanism of radionuclide release from the MWF alloys.

Abraham, D. P.; Peterson, J. J.; Katyal, H. K.; Keiser, D. D.; Hilton, B. A.

1999-12-14T23:59:59.000Z

44

Production of metal waste forms from spent fuel treatment  

SciTech Connect (OSTI)

Treatment of spent nuclear fuel at Argonne National Laboratory consists of a pyroprocessing scheme in which the development of suitable waste forms is being advanced. Of the two waste forms being proposed, metal and mineral, the production of the metal waste form utilizes induction melting to stabilize the waste product. Alloying of metallic nuclear materials by induction melting has long been an Argonne strength and thus, the transition to metallic waste processing seems compatible. A test program is being initiated to coalesce the production of the metal waste forms with current induction melting capabilities.

Westphal, B.R.; Keiser, D.D.; Rigg, R.H.; Laug, D.V.

1995-02-01T23:59:59.000Z

45

Shale Rocks as Nuclear Waste Repositories: Hydrothermal Reactions with Glass, Ceramic and Spent Fuel Waste Forms  

Science Journals Connector (OSTI)

The objectives of various contributions from this laboratory have been to simulate “worst case” situations, given a proposed choice of waste form, repository rock, and waste loading/waste age. The “worst case”...

W. Phelps Freeborn; Michael Zolensky…

1980-01-01T23:59:59.000Z

46

Waste Treatment Facility Passes Federal Inspection, Completes Final  

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

Waste Treatment Facility Passes Federal Inspection, Completes Final Waste Treatment Facility Passes Federal Inspection, Completes Final Milestone, Begins Startup Waste Treatment Facility Passes Federal Inspection, Completes Final Milestone, Begins Startup April 23, 2012 - 12:00pm Addthis Media Contact Erik Simpson, 208-390-9464 Danielle Miller, 208-526-5709 The Idaho site today initiated the controlled, phased startup of a new waste treatment facility scheduled to begin treating 900,000 gallons of radioactive liquid waste stored in underground tanks at a former Cold War spent nuclear fuel reprocessing facility next month. A U.S. Department of Energy (DOE) operational readiness review team (made up of Subject Matter Experts across the country) in early April identified a dozen issues for the cleanup contractor CH2M-WG Idaho, LLC (CWI) to

47

Microstructural characterization of halite inclusions in a surrogate glass bonded ceramic waste form  

SciTech Connect (OSTI)

A glass-bonded ceramic waste form is being developed to immobilize high-level chloride waste salts generated during the conditioning of spent sodium-bonded nuclear fuel for disposal. The waste salt is loaded into zeolite cavities, mixed with a borosilicate glass, and consolidated at 800--900 C by hot isostatic pressing. During this process, small amounts of halite are generated, whereas the zeolite converts to the mineral sodalite, which retains most of the waste salt. In this work, optical microscopy, scanning electron microscopy, and transmission electron microscopy2048e used to characterize the halite inclusions in the final waste form. The halite inclusions were detected within micron- to submicron-sized pores that form within the glass phase in the vicinity of the sodalite/glass interface. The chemical nature and distribution of the halite inclusions were determined. The particular microstructure of the halite inclusions has been related to the corrosion of the ceramic waste form.

Luo, J. S.; Zyryanov, V. N.; Ebert, W. L.

2000-05-12T23:59:59.000Z

48

Summary of INEL research on the iron-enriched basalt waste form  

SciTech Connect (OSTI)

This report summarizes the knowledge base on the iron-enriched basalt (IEB) waste form developed at the Idaho National Engineering Laboratory (INEL) during 1979--1982. The results presented discuss the applicability of IEB in converting retrieved transuranic (TRU) waste from INEL`s Radioactive Waste Management Complex (RWMC) into a vitreous/ceramic (glassy/rock) stable waste form suitable for permanent disposal in an appropriate repository, such as the Waste Isolation Pilot Plant (WIPP) in New Mexico. Borosilicate glass (BSG), the approved high-level waste form, appears unsuited for this application. Melting the average waste-soil mix from the RWMC produces the IEB composition and attempting to convert IEB to the BSG composition would require additions of substantial B{sub 2}0{sub 3}, Na, and SiO{sub 2} (glass frit). IEB requires processing temperatures of 1400 to 1600{degrees}C, depending upon the waste composition. Production of the IEB waste form, using Joule heated melters, has proved difficult in the past because of electrode and refractory corrosion problems associated with the high temperature melts. Higher temperature electric melters (arc and plasma) are available to produce this final waste form. Past research focused on extensive slag property measurements, waste form leachability tests, mechanical, composition, and microstructure evaluations, as well as a host of experiments to improve production of the waste form. Past INEL studies indicated that the IEB glass-ceramic is a material that will accommodate and stabilize a wide range of heterogeneous waste materials, including long lived radionuclides and scrap metals, while maintaining a superior level of chemical and physical performance characteristics. Controlled cooling of the molten IEB and subsequent heat treatment will produce a glass-ceramic waste form with superior leach resistance.

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

1992-01-01T23:59:59.000Z

49

Summary of INEL research on the iron-enriched basalt waste form  

SciTech Connect (OSTI)

This report summarizes the knowledge base on the iron-enriched basalt (IEB) waste form developed at the Idaho National Engineering Laboratory (INEL) during 1979--1982. The results presented discuss the applicability of IEB in converting retrieved transuranic (TRU) waste from INEL's Radioactive Waste Management Complex (RWMC) into a vitreous/ceramic (glassy/rock) stable waste form suitable for permanent disposal in an appropriate repository, such as the Waste Isolation Pilot Plant (WIPP) in New Mexico. Borosilicate glass (BSG), the approved high-level waste form, appears unsuited for this application. Melting the average waste-soil mix from the RWMC produces the IEB composition and attempting to convert IEB to the BSG composition would require additions of substantial B{sub 2}0{sub 3}, Na, and SiO{sub 2} (glass frit). IEB requires processing temperatures of 1400 to 1600{degrees}C, depending upon the waste composition. Production of the IEB waste form, using Joule heated melters, has proved difficult in the past because of electrode and refractory corrosion problems associated with the high temperature melts. Higher temperature electric melters (arc and plasma) are available to produce this final waste form. Past research focused on extensive slag property measurements, waste form leachability tests, mechanical, composition, and microstructure evaluations, as well as a host of experiments to improve production of the waste form. Past INEL studies indicated that the IEB glass-ceramic is a material that will accommodate and stabilize a wide range of heterogeneous waste materials, including long lived radionuclides and scrap metals, while maintaining a superior level of chemical and physical performance characteristics. Controlled cooling of the molten IEB and subsequent heat treatment will produce a glass-ceramic waste form with superior leach resistance.

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

1992-01-01T23:59:59.000Z

50

Accelerated chemical aging of crystalline nuclear waste forms  

Science Journals Connector (OSTI)

Nuclear waste disposal is a significant technological issue, and the solution of this problem (or lack thereof) will ultimately determine whether nuclear energy is deemed environmentally friendly, despite significantly lower carbon emissions than fossil fuel energy sources. A critical component of any waste disposal strategy is the selection of the waste form that is tasked with preventing radionuclides from entering the environment. The design of robust nuclear waste forms requires consideration of several criteria, including: radiation tolerance, geological interaction and chemical durability; all of these criteria ensure that the radionuclides do not escape from the waste form. Over the past 30 years, there have been numerous and thorough studies of these criteria on candidate waste forms, including radiation damage and leaching. However, most of these efforts have focused on the performance of the candidate waste form at t = 0, with far less attention paid to the phase stability, and subsequent durability, of candidate waste forms during the course of daughter product formation; that is, the chemical aging of the material. Systematic understanding of phase evolution as a function of chemistry is important for predictions of waste form performance as well as informing waste form design. In this paper, we highlight the research challenges associated with understanding waste form stability when attempting to systematically study the effects of dynamic composition variation due to in situ radionuclide daughter production formation.

C.R. Stanek; B.P. Uberuaga; B.L. Scott; R.K. Feller; N.A. Marks

2012-01-01T23:59:59.000Z

51

Support for DOE program in mineral waste-form development  

SciTech Connect (OSTI)

This research investigation relates to sintered simulation ceramic waste forms of the generic SYNROC compositional type. Though they have been formulated with simulated wastes only, they serve as prototypes for potential hot, processed, crystalline waste forms whose combined thermodynamic stability and physical integrity are considered to render them capable of long-term imobilization of high-level radwastes under deep geologic disposal conditions. The problems involved are nontrivial, largely because of the very complex nature of the radwastes: a typical waste stream would contain more than 31 cation species. When the stabilizing matrix constituents are included, the final batch composition must successfully account (and find substitutional homes for some 35 different cation species. One of the important objectives of this study thus has been to develop a computer-based method for simulating these complex ion substitutions, and for calculating the resultant phase demands and batch formulations. Primary goals of the study have been (1) use of that computer simulation capability to incorporate rationally the radwaste ions from a specific waste stream (PW-7a) into the available SYNROC lattice sites and (2) utilization of existing ceramic processing and sintering methodologies to assure (and to understand) the attainment of high density, fine microstructure, full phase development and other features of the sintered product which are known to relate directly to its integrity and leach resistance. Though improved resistance to leaching has been a continuing goal, time and budget constraints have precluded initiation of any leachability studies of these new compositions during this contract period. 27 references, 15 figures, 6 tables.

Palmour, H. III; Hare, T.M.; Russ, J.C.; Batchelor, A.D.; Paisley, M.J.; Freed, L.E.

1982-09-01T23:59:59.000Z

52

CSNF WASTE FORM DEGRADATION: SUMMARY ABSTRACTION  

SciTech Connect (OSTI)

The purpose of this model report is to describe the development and validation of models that can be used to calculate the release of radionuclides from commercial spent nuclear fuel (CSNF) following a hypothetical breach of the waste package and fuel cladding in the repository. The purpose also includes describing the uncertainties associated with modeling the radionuclide release for the range of CSNF types, exposure conditions, and durations for which the radionuclide release models are to be applied. This document was developed in accordance with Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 169944]). This document considers radionuclides to be released from CSNF when they are available for mobilization by gas-phase mass transport, or by dissolution or colloid formation in water that may contact the fuel. Because other reports address limitations on the dissolved and colloidal radionuclide concentrations (BSC 2004 [DIRS 169944], Table 2-1), this report does not address processes that control the extent to which the radionuclides released from CSNF are mobilized and transported away from the fuel either in the gas phase or in the aqueous phase as dissolved and colloidal species. The scope is limited to consideration of degradation of the CSNF rods following an initial breach of the cladding. It considers features of CSNF that limit the availability of individual radionuclides for release into the gaseous or aqueous phases that may contact the fuel and the processes and events expected to degrade these CSNF features. In short, the purpose is to describe the characteristics of breached fuel rods and the degradation processes expected to influence radionuclide release.

J.C. CUNNANE

2004-08-31T23:59:59.000Z

53

Waste package/repository impact study: Final report  

SciTech Connect (OSTI)

The Waste Package/Repository Impact Study was conducted to evaluate the feasibility of using the current reference salt waste package in the salt repository conceptual design. All elements of the repository that may impact waste package parameters, i.e., (size, weight, heat load) were evaluated. The repository elements considered included waste hoist feasibility, transporter and emplacement machine feasibility, subsurface entry dimensions, feasibility of emplacement configuration, and temperature limits. The evaluations are discussed in detail with supplemental technical data included in Appendices to this report, as appropriate. Results and conclusions of the evaluations are discussed in light of the acceptability of the current reference waste package as the basis for salt conceptual design. Finally, recommendations are made relative to the salt project position on the application of the reference waste package as a basis for future design activities. 31 refs., 11 figs., 11 tabs.

Not Available

1985-09-01T23:59:59.000Z

54

Waste Management's LNG Truck Fleet: Final Results  

SciTech Connect (OSTI)

Waste Management, Inc., began operating a fleet of heavy-duty LNG refuse trucks at its Washington, Pennsylvania, facility. The objective of the project was to provide transportation professionals with quantitative, unbiased information on the cost, maintenance, operational, and emissions characteristics of LNG as one alternative to conventional diesel for heavy-duty trucking applications.

Chandler, K. [Battelle (US); Norton, P. [National Renewable Energy Laboratory (US); Clark, N. [West Virginia University (US)

2001-01-25T23:59:59.000Z

55

Energy utilization: municipal waste incineration. Final report  

SciTech Connect (OSTI)

An assessment is made of the technical and economical feasibility of converting municipal waste into useful and useable energy. The concept presented involves retrofitting an existing municipal incinerator with the systems and equipment necessary to produce process steam and electric power. The concept is economically attractive since the cost of necessary waste heat recovery equipment is usually a comparatively small percentage of the cost of the original incinerator installation. Technical data obtained from presently operating incinerators designed specifically for generating energy, documents the technical feasibility and stipulates certain design constraints. The investigation includes a cost summary; description of process and facilities; conceptual design; economic analysis; derivation of costs; itemized estimated costs; design and construction schedule; and some drawings.

LaBeck, M.F.

1981-03-27T23:59:59.000Z

56

Iron Oxide Waste Form for Stabilizing 99Tc. | EMSL  

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

Stabilizing 99Tc. Iron Oxide Waste Form for Stabilizing 99Tc. Abstract: Crystals of goethite were synthesized with reduced technetium 99Tc(IV) incorporated within the solid...

57

Waste Form Degradation Model Integration for Engineered Materials Performance  

Broader source: Energy.gov [DOE]

The collaborative approach to the glass and metallic waste form degradation modeling activities includes process model development (including first-principles approaches) and model integration—both...

58

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

SciTech Connect (OSTI)

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). The pretreatment facility will have the capacity to separate all of the tank wastes into the HLW and LAW fractions, and the HLW Vitrification Facility will have the capacity to vitrify all of the HLW. However, a second immobilization facility will be needed for the expected volume of LAW requiring immobilization. A number of alternatives, including Cast Stone—a cementitious waste form—are being considered to provide the additional LAW immobilization capacity.

Westsik, Joseph H.; Serne, R. Jeffrey; Pierce, Eric M.; Cozzi, Alex; Chung, Chul-Woo; Swanberg, David J.

2013-05-31T23:59:59.000Z

59

Forms of Al in Hanford Tank Waste  

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

Actual Waste Testing Actual Waste Testing Lanée Snow Sandra Fiskum Rick Shimskey Reid Peterson 4/9/09 2 Tested > 75% of sludge waste types Sludge Sources Bi-Phosphate waste Redox Purex Cladding TBP FeCN sludge Redox Cladding Zirc Cladding Purex waste Misc NA 4/9/09 3 Tested > 75% of saltcake waste types Saltcake fractions Bi-phosphate saltcake S A B R NA Tested 8 groups of tank waste types Group ID Type Al Cr PO 4 3- Oxalate Sulfate Fluoride 1 Bi Phosphate sludge 3% 3% 21% 2% 6% 12% 2 Bi Phosphate saltcake (BY, T) 18% 25% 36% 36% 43% 36% 3 PUREX Cladding Waste sludge 12% 1% 3% 1% 1% 3% 4 REDOX Cladding Waste sludge 8% 1% 0% 0% 0% 2% 5 REDOX sludge 26% 8% 1% 3% 1% 2% 6 S - Saltcake (S) 11% 38% 12% 24% 14% 3% 7 TBP Waste sludge 1% 1% 8% 0% 2% 1% 8 FeCN sludge 2% 1% 4% 1% 1% 1% *Percentages reflect % of total inventory of species in the tank farm. *Discussion will focus on those that make up the largest fraction of the Al

60

Secondary waste form testing : ceramicrete phosphate bonded ceramics.  

SciTech Connect (OSTI)

The cleanup activities of the Hanford tank wastes require stabilization and solidification of the secondary waste streams generated from the processing of the tank wastes. The treatment of these tank wastes to produce glass waste forms will generate secondary wastes, including routine solid wastes and liquid process effluents. Liquid wastes may include process condensates and scrubber/off-gas treatment liquids from the thermal waste treatment. The current baseline for solidification of the secondary wastes is a cement-based waste form. However, alternative secondary waste forms are being considered. In this regard, Ceramicrete technology, developed at Argonne National Laboratory, is being explored as an option to solidify and stabilize the secondary wastes. The Ceramicrete process has been demonstrated on four secondary waste formulations: baseline, cluster 1, cluster 2, and mixed waste streams. Based on the recipes provided by Pacific Northwest National Laboratory, the four waste simulants were prepared in-house. Waste forms were fabricated with three filler materials: Class C fly ash, CaSiO{sub 3}, and Class C fly ash + slag. Optimum waste loadings were as high as 20 wt.% for the fly ash and CaSiO{sub 3}, and 15 wt.% for fly ash + slag filler. Waste forms for physical characterizations were fabricated with no additives, hazardous contaminants, and radionuclide surrogates. Physical property characterizations (density, compressive strength, and 90-day water immersion test) showed that the waste forms were stable and durable. Compressive strengths were >2,500 psi, and the strengths remained high after the 90-day water immersion test. Fly ash and CaSiO{sub 3} filler waste forms appeared to be superior to the waste forms with fly ash + slag as a filler. Waste form weight loss was {approx}5-14 wt.% over the 90-day immersion test. The majority of the weight loss occurred during the initial phase of the immersion test, indicative of washing off of residual unreacted binder components from the waste form surface. Waste forms for ANS 16.1 leach testing contained appropriate amounts of rhenium and iodine as radionuclide surrogates, along with the additives silver-loaded zeolite and tin chloride. The leachability index for Re was found to range from 7.9 to 9.0 for all the samples evaluated. Iodine was below detection limit (5 ppb) for all the leachate samples. Further, leaching of sodium was low, as indicated by the leachability index ranging from 7.6-10.4, indicative of chemical binding of the various chemical species. Target leachability indices for Re, I, and Na were 9, 11, and 6, respectively. Degradation was observed in some of the samples post 90-day ANS 16.1 tests. Toxicity characteristic leaching procedure (TCLP) results showed that all the hazardous contaminants were contained in the waste, and the hazardous metal concentrations were below the Universal Treatment Standard limits. Preliminary scale-up (2-gal waste forms) was conducted to demonstrate the scalability of the Ceramicrete process. Use of minimal amounts of boric acid as a set retarder was used to control the working time for the slurry. Flexibility in treating waste streams with wide ranging compositional make-ups and ease of process scale-up are attractive attributes of Ceramicrete technology.

Singh, D.; Ganga, R.; Gaviria, J.; Yusufoglu, Y. (Nuclear Engineering Division); ( ES)

2011-06-21T23:59:59.000Z

Note: This page contains sample records for the topic "final waste form" 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

EVALUATION OF THE FINAL REPORT: WASTE PACKAGE MATERIALS PERFORMANCE PEER  

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

EVALUATION OF THE FINAL REPORT: WASTE EVALUATION OF THE FINAL REPORT: WASTE PACKAGE MATERIALS PERFORMANCE PEER REVIEW PANEL B00000000-01717-5700-00005 REV 00 August 2002 This document is not an official copy and is for informational purposes only. QA: QA B00000000-01717-5700-00005 REV 00 August 2002 Evaluation of the Final Report: Waste Package Materials Performance Peer Review Panel Prepared by: Jack N. Bailey, Jack D. Cloud, Thomas E. Rodgers, and Tammy S.E. Summers Prepared for: U.S. Department of Energy Yucca Mountain Site Characterization Office P.O. Box 364629 North Las Vegas, Nevada 89036-8629 Prepared by: Bechtel SAIC Company, LLC 1180 Town Center Drive Las Vegas, Nevada 89144 Under Contract Number DE-AC28-01RW12101 Disclaimer Signature Page Change History Acknowledgments

62

Waste treatment facility passes federal inspection, completes final  

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

23, 2012 23, 2012 Media Contact: Danielle Miller, 208-526-5709 Erik Simpson, 208-390-9464 Waste treatment facility passes federal inspection, completes final milestone, begins startup The Idaho site today initiated the controlled, phased startup of a new waste treatment facility scheduled to begin treating 900,000 gallons of radioactive liquid waste stored in underground tanks at a former Cold War spent nuclear fuel reprocessing facility next month. An exterior view of the Integrated Waste Treatment Unit A U.S. Department of Energy (DOE) operational readiness review team (made up of Subject Matter Experts across the country) in early April identified a dozen issues for the cleanup contractor CH2M-WG Idaho, LLC (CWI) to resolve before the 53,000-square-foot Integrated Waste Treatment Unit

63

Final environmental impact statement. Waste Isolation Pilot Plant  

SciTech Connect (OSTI)

This volume contains the appendices for the Final Environmental Impact Statement for the Waste Isolation Pilot Plant (WIPP). Alternative geologic environs are considered. Salt, crystalline rock, argillaceous rock, and tuff are discussed. Studies on alternate geologic regions for the siting of WIPP are reviewed. President Carter's message to Congress on the management of radioactive wastes and the findings and recommendations of the interagency review group on nuclear waste management are included. Selection criteria for the WIPP site including geologic, hydrologic, tectonic, physicochemical compatability, and socio-economic factors are presented. A description of the waste types and the waste processing procedures are given. Methods used to calculate radiation doses from radionuclide releases during operation are presented. A complete description of the Los Medanos site, including archaeological and historic aspects is included. Environmental monitoring programs and long-term safety analysis program are described. (DMC)

Not Available

1980-10-01T23:59:59.000Z

64

A mathematical model to predict leaching of hazardous inorganic wastes from solidified/stabilized waste forms  

E-Print Network [OSTI]

A MATHEMATICAL MODEL TO PREDICT LEACHING OF HAZARDOUS INORGANIC WASTES FROM SOLIDIFIED/STABILIZED WASTE FORMS A Thesis by KRISHAN SABHARWAL Submitted to the Office of Graduate Studies of Texas AkM University in partial fulfillment...A MATHEMATICAL MODEL TO PREDICT LEACHING OF HAZARDOUS INORGANIC WASTES FROM SOLIDIFIED/STABILIZED WASTE FORMS A Thesis by KRISHAN SABHARWAL Submitted to the Office of Graduate Studies of Texas AkM University in partial fulfillment...

Sabharwal, Krishan

2012-06-07T23:59:59.000Z

65

DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION  

SciTech Connect (OSTI)

Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an ''upper-limit'' (i.e., instantaneous degradation) model for use in the TSPA-LA model. ''Best-estimate'' models for the degradation of the fuels in each of the DSNF groups are discussed to provide a basis for selecting the upper limit model for use in the TSPA-LA model. The instantaneous degradation model is chosen for use in the TSPA-LA model because the available information shows that the degradation rate of the N Reactor fuel (which constitutes most of the DSNF inventory) is very high and because the available qualified information is insufficient to justify use of a less conservative approach. The commercial spent nuclear fuel model will be used for naval spent nuclear fuel because it has been shown to be conservative for representing naval spent nuclear fuel.

J. CUNNANE

2004-11-19T23:59:59.000Z

66

EIS-0287: Idaho High-Level Waste and Facilities Disposition Final...  

Office of Environmental Management (EM)

: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002) EIS-0287: Idaho High-Level Waste and Facilities Disposition...

67

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

SciTech Connect (OSTI)

To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions has initiated secondary-waste-form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is conducting tests on four candidate waste forms to evaluate their ability to meet potential waste acceptance criteria for immobilized secondary wastes that would be placed in the IDF. All three waste forms demonstrated compressive strengths above the minimum 3.45 MPa (500 psi) set as a target for cement-based waste forms. Further, none of the waste forms showed any significant degradation in compressive strength after undergoing thermal cycling (30 cycles in a 10 day period) between -40 C and 60 C or water immersion for 90 days. The three leach test methods are intended to measure the diffusion rates of contaminants from the waste forms. Results are reported in terms of diffusion coefficients and a leachability index (LI) calculated based on the diffusion coefficients. A smaller diffusion coefficient and a larger LI are desired. The NRC, in its Waste Form Technical Position (NRC 1991), provides recommendations and guidance regarding methods to demonstrate waste stability for land disposal of radioactive waste. Included is a recommendation to conduct leach tests using the ANS 16.1 method. The resulting leachability index (LI) should be greater than 6.0. For Hanford secondary wastes, the LI > 6.0 criterion applies to sodium leached from the waste form. For technetium and iodine, higher targets of LI > 9 for Tc and LI > 11 for iodine have been set based on early waste-disposal risk and performance assessment analyses. The results of these three leach tests conducted for a total time between 11days (ASTM C1308) to 90 days (ANS 16.1) showed: (1) Technetium diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that all the waste forms had leachability indices better than the target LI > 9 for technetium; (2) Rhenium diffusivity: Cast Stone 2M specimens, when tested using EPA 1315 protocol, had leachability indices better than the target LI > 9 for technetium based on rhenium as a surrogate for technetium. All other waste forms tested by ANSI/ANS 16.1, ASTM C1308, and EPA 1315 test methods had leachability indices that were below the target LI > 9 for Tc based on rhenium release. These studies indicated that use of Re(VII) as a surrogate for 99Tc(VII) in low temperature secondary waste forms containing reductants will provide overestimated diffusivity values for 99Tc. Therefore, it is not appropriate to use Re as a surrogate 99Tc in future low temperature waste form studies. (3) Iodine diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that the three waste forms had leachability indices that were below the target LI > 11 for iodine. Therefore, it may be necessary to use a more effective sequestering material than silver zeolite used in two of the waste forms (Ceramicrete and DuraLith); (4) Sodium diffusivity: All the waste form specimens tested by the three leach methods (ANSI/ANS 16.1, ASTM C1308, and EPA 1315) exceeded the target LI value of 6; (5) All three leach methods (ANS 16.1, ASTM C1308 and EPA 1315) provided similar 99Tc diffusivity values for both short-time transient diffusivity effects as well as long-term ({approx}90 days) steady diffusivity from each of the three tested waste forms (Cast Stone 2M, Ceramicrete and DuraLith). Therefore, any one of the three methods can be used to determine the contaminant diffusivities from a selected waste form.

Mattigod, Shas V.; Westsik, Joseph H.; Chung, Chul-Woo; Lindberg, Michael J.; Parker, Kent E.

2011-08-12T23:59:59.000Z

68

Application of PCT to the EBR II ceramic waste form.  

SciTech Connect (OSTI)

We are evaluating the use of the Product Consistency Test (PCT) developed to monitor the consistency of borosilicate glass waste forms for application to the multiphase ceramic waste form (CWF) that will be used to immobilize waste salts generated during the electrometallurgical conditioning of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor No. 2 (EBR II). The CWF is a multiphase waste form comprised of about 70% sodalite, 25% borosilicate glass binder, and small amounts of halite and oxide inclusions. It must be qualified for disposal as a non-standard high-level waste (HLW) form. One of the requirements in the DOE Waste Acceptance System Requirements Document (WASRD) for HLW waste forms is that the consistency of the waste forms be monitored.[1] Use of the PCT is being considered for the CWF because of the similarities of the dissolution behaviors of both the sodalite and glass binder phases in the CWF to borosilicate HLW glasses. This paper provides (1) a summary of the approach taken in selecting a consistency test for CWF production and (2) results of tests conducted to measure the precision and sensitivity of the PCT conducted with simulated CWF.

Ebert, W. L.; Lewis, M. A.; Johnson, S. G.

2002-01-10T23:59:59.000Z

69

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

SciTech Connect (OSTI)

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

COZZI, ALEX

2004-02-18T23:59:59.000Z

70

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

SciTech Connect (OSTI)

Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomic Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration project will provide additional options for fission product immobilization and waste management associated the electrochemical/pyrometallurgical processing of used nuclear fuel.

S.M. Frank

2011-09-01T23:59:59.000Z

71

Method for forming microspheres for encapsulation of nuclear waste  

DOE Patents [OSTI]

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

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

1984-01-01T23:59:59.000Z

72

Final environmental impact statement. Waste Isolation Pilot Plant  

SciTech Connect (OSTI)

In accordance with the National Environmental Policy Act (NEPA) of 1969, the US Department of Energy (DOE) has prepared this document as environmental input to future decisions regarding the Waste Isolation Pilot Plant (WIPP), which would include the disposal of transuranic waste, as currently authorized. The alternatives covered in this document are the following: (1) Continue storing transuranic (TRU) waste at the Idaho National Engineering Laboratory (INEL) as it is now or with improved confinement. (2) Proceed with WIPP at the Los Medanos site in southeastern New Mexico, as currently authorized. (3) Dispose of TRU waste in the first available repository for high-level waste. The Los Medanos site would be investigated for its potential suitability as a candidate site. This is administration policy and is the alternative preferred by the DOE. (4) Delay the WIPP to allow other candidate sites to be evaluated for TRU-waste disposal. This environmental impact statement is arranged in the following manner: Chapter 1 is an overall summary of the analysis contained in the document. Chapters 2 and 4 set forth the objectives of the national waste-management program and analyze the full spectrum of reasonable alternatives for meeting these objectives, including the WIPP. Chapter 5 presents the interim waste-acceptance criteria and waste-form alternatives for the WIPP. Chapters 6 through 13 provide a detailed description and environmental analysis of the WIPP repository and its site. Chapter 14 describes the permits and approvals necessary for the WIPP and the interactions that have taken place with Federal, State, and local authorities, and with the general public in connection with the repository. Chapter 15 analyzes the many comments received on the DEIS and tells what has been done in this FEIS in response. The appendices contain data and discussions in support of the material in the text.

Not Available

1980-10-01T23:59:59.000Z

73

West Valley Demonstration Project Waste Management Final Environmental Impact Statement  

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

WEST VALLEY DEMONSTRATION PROJECT WEST VALLEY DEMONSTRATION PROJECT WASTE MANAGEMENT ENVIRONMENTAL IMPACT STATEMENT FINAL SUMMARY December 2003 Prepared by: U.S. Department of Energy West Valley Area Office West Valley, NY DOE/EIS - 0337F For general questions or to request a copy of this EIS, please contact: DANIEL W. SULLIVAN, DOCUMENT MANAGER DOE WEST VALLEY AREA OFFICE 10282 Rock Springs Road WEST VALLEY, NY 14171-0191 1-800-633-5280 COVER SHEET Lead Agency: U.S. Department of Energy Title: Final West Valley Demonstration Project Waste Management Environmental Impact Statement, Cattaraugus County, West Valley, New York. Contact: For further information about this Environmental Impact Statement, contact: For general information on the Department of Energy's process for implementing the National

74

Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams  

SciTech Connect (OSTI)

In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste stream options in terms of waste loading and/or decay time required before treatment. For Option 1, glass ceramics show an increase in waste loading of 15 mass % and reduction in decay time of 24 years. Decay times of {approx}50 years or longer are close to the expected age of the fuel that will be reprocessed when the modified open or closed fuel cycle is expected to be put into action. Option 2 shows a 2x to 2.5x increase in waste loading with decay times of only 45 years. Note that for Option 2 glass, the required decay time before treatment is only 35 years because of the waste loading limits related to the solubility of MoO{sub 3} in glass. If glass was evaluated for similar waste loadings as those achieved in Option 2 glass ceramics, the decay time would be significantly longer than 45 years. These glass ceramics are not optimized, but already they show the potential to dramatically reduce the amount of waste generated while still utilizing the proven processing technology used for glass production.

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

2010-09-23T23:59:59.000Z

75

Chapter 13 - Actinide host phases as radioactive waste forms  

Science Journals Connector (OSTI)

Publisher Summary An effective strategy for dealing with high-level waste is to partition the short-lived fission product elements from the long-lived actinides, creating separate waste streams. Once there are two waste streams, the properties and durability of the waste form can be designed to a level appropriate to the toxicity and time required for isolation from the environment. With such a strategy the fission product elements may be incorporated into a borosilicate glass and the actinides into more durable crystalline ceramics. Although special glass compositions may be developed for actinide incorporation, their long-term durability is less easily assured, particularly on the time scales required for actinide immobilization and confinement. The final selection of any waste form should depend on its ability to incorporate the radionuclides of interest, its chemical durability, response to a radiation-field, and physical properties as well as the time required for isolation to protect the environment. There are three significant types of actinide-containing materials generated by the nuclear fuel cycle that contain high levels of radioactivity: 1.) spent nuclear fuel (SNF) related to the production of fissile material for weapons, 2.) SNF from commercial nuclear reactors, 3.) liquid high-level waste (HLW) derived during the reprocessing of SNF [1]. Unreacted fuel constituents (235,238U) make up approximately 96% of total mass of SNF. A major fraction of activity of SNF comes from fission product (FP) elements with mass numbers from 85 to 106 and from 125–147 (Kr, Sr, Y, Zr, Tc, Ru, Y, Sb, Cs, Ba, Ce, Pm, etc.), unreacted fuel (U), “minor” actinides (Np, Pu, Am, Cm), and activated products (H, C, Al, Na, Mn, Fe, Co). \\{FPs\\} consist of about 200 isotopes of approximately 40 elements from Zn to Gd. The yield of individual radionuclides ranges between 104 % to several percent (a yield of 1 % corresponds to production of 1 atom of daughter isotope per 100 events of nuclear decay of 235U or 239Pu). The fraction of individual radionuclides in SNF varies depending on the type of reactor, burn-up and cooling time. From point of view of radiobiological risk the following groups of radionuclides are important:u• Short-lived \\{FPs\\} which are almost completely decayed to stable isotopes after a cooling of SNF for some tens of years: Rb, Y, Mo, Ru, Rh, Ag, Sb, Te, Xe, Ba, La, Ce, Pr, Nd, Pm. Their amount in total is 26 kg per metric tone (MT) of SNF or 65 wt.% of the total \\{FPs\\} amount; • \\{FPs\\} with high specific activity: mainly 90Sr and 137Cs; their total content is up to 6 kg per 1 MT of SNF (about 15 wt.% of total FPs); • Long-lived \\{FPs\\} with low specific activity: Zr, Tc, Pd, Sn, I (about 8 kg per 1 MT of SNF or about 20 wt.% of total FPs); • Actinides (Np, Pu, Am, Cm) and their daughter products which are less than 1 wt.% and dominated by Pu; • Unreacted constituents: 238U - 98.9 wt.% and 235U -1.1 wt.% of total.

Sergey V. Yudintsev; Sergey V. Stefanovsky; Rodney C. Ewing

2007-01-01T23:59:59.000Z

76

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

SciTech Connect (OSTI)

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

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

1998-05-01T23:59:59.000Z

77

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

SciTech Connect (OSTI)

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

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

1982-09-01T23:59:59.000Z

78

Characteristics of metal waste forms containing technetium and uranium  

SciTech Connect (OSTI)

2 prototype alloys: RAW-1(Tc) and RAW-2(UTc) suitable for a wide range of waste stream compositions are being evaluated to support development of a waste form degradation model that can be used to calculate radionuclide source terms for a range of waste form compositions and disposal environments. Tests and analyses to support formulation of waste forms and development of the degradation model include detailed characterizations of the constituent phases using SEM/EDS and TEM, electrochemical tests to quantify the oxidation behavior and kinetics of the individual and coupled phases under a wide range of environmental conditions, and corrosion tests to measure the gross release kinetics of radionuclides under aggressive test conditions.

Fortner, J.A.; Kropf, A.J.; Ebert, W.L. [Argonne National Laboratory, Argonne, IL 60439 (United States)

2013-07-01T23:59:59.000Z

79

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

SciTech Connect (OSTI)

The formulation, synthesis, characterization and hydration kinetics of hydroceramic waste forms designed as potential hosts for existing INEEL calcine high-level wastes have been established as functions of temperature and processing time. Initial experimentations were conducted with several aluminosilicate pozzolanic materials, ranging from fly ash obtained from various power generating coal and other combustion industries to reactive alumina, natural clays and ground bottled glass powders. The final selection criteria were based on the ease of processing, excellent physical properties and chemical durability (low-leaching) determined from the PCT test produced in hydroceramic. The formulation contains vermiculite, Sr(NO32), CsC1, NaOH, thermally altered (calcined natural clay) and INEEL simulated calcine high-level nuclear wastes and 30 weight percent of fluorinel blend calcine and zirconia calcine. Syntheses were carried out at 75-200 degree C at autogeneous water pressure (100% relative humidity) at various time intervals. The resulting monolithic compact products were hard and resisted breaking when dropped from a 5 ft height. Hydroceramic host mixed with fluorinel blend calcine and processed at 75 degree C crumbled into rice hull-side grains or developed scaly flakes. However, the samples equally possessed the same chemical durability as their unbroken counterparts. Phase identification by XRD revealed that hydroceramic host crystallized type zeolite at 75-150 degree C and NaP1 at 175-200 degree C in addition to the presence of quartz phase originating from the clay reactant. Hydroceramic host mixed with either fluorinel blend calcine or zirconia calcine crystallized type A zeolite at 75-95 degree C, formed a mixture of type A zeolite and hydroxysodalite at 125-150 degree C and hydroxysodalite at 175-200 degree C. Quartz, calcium fluoride and zirconia phases from the clay reactant and the two calcine wastes were also detected. The PCT test solution conductivity, pH and analytical concentration measured as a function of time decrease exponentially. In some cases nitrate, sulfate, chloride and fluoride ion concentrations increased with time and processing temperature with respect to the reference sample. The increasing concentration of these ions was due to the lack of formation of crystalline phases that can incorporate them in their structures, especially cancrinite. Another plausible explanations for their increase was due to the continuous withdrawal of cations with time, for example sodium to form zeolites, thereby increase their concentrations.

Barry Scheetz; Johnson Olanrewaju

2001-10-15T23:59:59.000Z

80

DOE-EA-0179; Waste Form Selection for Savannah River Plant High-Level Waste  

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

48326 (F.R.) 48326 (F.R.) NOTICES DEPARTMENT OF ENERGY Compliance With the National Environmental Policy Act Proposed Finding of No Significant Impact, Selection of Borosilicate Glass as the Defense Waste Processing Facility Waste Form for High -Level Radioactive Wastes Savanah River Plant, Aiken, South Carolina Thursday, July 29, 1982 *32778 AGENCY: Energy Department. ACTION: Notice. SUMMARY: The Department of Energy (DOE) has prepared an environmental assessment (DOE/EA- 0179) on the proposed selection of borosilicate glass as the Defense Waste Processing Facility (DWPF) waste form for the immobilization of the high -level radioactive wastes generated and stored at the DOE Savannah River Plant (SRP), Aiken, South Carolina. DOE recently decided to immobilize

Note: This page contains sample records for the topic "final waste form" 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

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

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

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

82

Waste Form Release Data Package for the 2001 Immobilized Low-Activity Waste Performance Assessment  

SciTech Connect (OSTI)

This data package documents the experimentally derived input data on the representative waste glasses LAWABP1 and HLP-31 that will be used for simulations of the immobilized lowactivity waste disposal system with the Subsurface Transport Over Reactive Multiphases (STORM) code. The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in March of 2001. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali-H ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow and vapor hydration experiments were used for accelerated weathering or aging of the glasses. The majority of the thermodynamic data were extracted from the thermodynamic database package shipped with the geochemical code EQ3/6. However, several secondary reaction products identified from laboratory tests with prototypical LAW glasses were not included in this database, nor are the thermodynamic data available in the open literature. One of these phases, herschelite, was determined to have a potentially significant impact on the release calculations and so a solubility product was estimated using a polymer structure model developed for zeolites. Although this data package is relatively complete, final selection of ILAW glass compositions has not been done by the waste treatment plant contractor. Consequently, revisions to this data package to address new ILAW glass formulations are to be regularly expected.

McGrail, B. Peter; Icenhower, Jonathan P.; Martin, Paul F.; Schaef, Herbert T.; O'Hara, Matthew J.; Rodriguez, Eugenio; Steele, Jackie L.

2001-02-01T23:59:59.000Z

83

Microsoft PowerPoint - Marcinowski - Waste Management (FINAL...  

Office of Environmental Management (EM)

Update on WIPP, Tank Waste and Other Waste Disposition Frank Marcinowski Deputy Assistant Secretary for Waste Management Office of Environmental Management EM SSAB Chairs Fall...

84

Reference Alloy Waste Form Fabrication and Initiation of Reducing Atmosphere and Reductive Additives Study on Alloy Waste Form Fabrication  

SciTech Connect (OSTI)

This report describes the fabrication of two reference alloy waste forms, RAW-1(Re) and RAW-(Tc) using an optimized loading and heating method. The composition of the alloy materials was based on a generalized formulation to process various proposed feed streams resulting from the processing of used fuel. Waste elements are introduced into molten steel during alloy fabrication and, upon solidification, become incorporated into durable iron-based intermetallic phases of the alloy waste form. The first alloy ingot contained surrogate (non-radioactive), transition-metal fission products with rhenium acting as a surrogate for technetium. The second alloy ingot contained the same components as the first ingot, but included radioactive Tc-99 instead of rhenium. Understanding technetium behavior in the waste form is of particular importance due the longevity of Tc-99 and its mobility in the biosphere in the oxide form. RAW-1(Re) and RAW-1(Tc) are currently being used as test specimens in the comprehensive testing program investigating the corrosion and radionuclide release mechanisms of the representative alloy waste form. Also described in this report is the experimental plan to study the effects of reducing atmospheres and reducing additives to the alloy material during fabrication in an attempt to maximize the oxide content of waste streams that can be accommodated in the alloy waste form. Activities described in the experimental plan will be performed in FY12. The first aspect of the experimental plan is to study oxide formation on the alloy by introducing O2 impurities in the melt cover gas or from added oxide impurities in the feed materials. Reducing atmospheres will then be introduced to the melt cover gas in an attempt to minimize oxide formation during alloy fabrication. The second phase of the experimental plan is to investigate melting parameters associated with alloy fabrication to allow the separation of slag and alloy components of the melt.

S.M. Frank; T.P. O'Holleran; P.A. Hahn

2011-09-01T23:59:59.000Z

85

Organic tanks safety program waste aging studies. Final report, Revision 1  

SciTech Connect (OSTI)

Uranium and plutonium production at the Hanford Site produced large quantities of radioactive byproducts and contaminated process chemicals that are stored in underground tanks awaiting treatment and disposal. Having been made strongly alkaline and then subjected to successive water evaporation campaigns to increase storage capacity, the wastes now exist in the physical forms of saltcakes, metal oxide sludges, and aqueous brine solutions. Tanks that contain organic process chemicals mixed with nitrate/nitrite salt wastes might be at risk for fuel-nitrate combustion accidents. This project started in fiscal year 1993 to provide information on the chemical fate of stored organic wastes. While historical records had identified the organic compounds originally purchased and potentially present in wastes, aging experiments were needed to identify the probable degradation products and evaluate the current hazard. The determination of the rates and pathways of degradation have facilitated prediction of how the hazard changes with time and altered storage conditions. Also, the work with aged simulated waste contributed to the development of analytical methods for characterizing actual wastes. Finally, the results for simulants provide a baseline for comparing and interpreting tank characterization data.

Camaioni, D.M.; Samuels, W.D.; Linehan, J.C. [and others

1998-09-01T23:59:59.000Z

86

Technetium Waste Form Development - Progress Report  

SciTech Connect (OSTI)

Analytical electron microscopy using SEM and TEM has been used to analyze a ~5 g. ingot with composition 71.3 wt% 316SS-5.3 wt% Zr-13.2 wt% Mo-4.0 wt% Rh-6.2 wt% Re prepared at the Idaho National Laboratory. Four phase fields have been identified two of which are lamellar eutectics, with a fifth possibly present. A Zr rich phase was found distributed as fine precipitate, ~10µm in diameter, often coating large cavities. A Mo-Fe-Re-Cr lamellar eutectic phase field appears as blocky regions ~30µm in diameter, surrounded by a Fe-Mo-Cr lamellar eutectic phase field, and that in turn is surrounded by a Zr-Fe-Rh-Mo-Ni phase field. The eutectic phase separation reactions are different. The Mo-Fe-Re-Cr lamellar eutectic appears a result of austenitic steel forming at lower volume fraction within an Mo-Fe-Re intermetallic phase, whereas the Fe-Mo-Cr lamellar eutectic may be a result of the same intermetallic phase forming within a ferritic steel phase. Cavitation may have arisen either as a result of bubbles, or from loss of equiaxed particles during specimen preparation.

Gelles, David S.; Ermi, Ruby M.; Buck, Edgar C.; Seffens, Rob J.; Chamberlin, Clyde E.

2009-01-07T23:59:59.000Z

87

Transuranic waste form characterization and data base. Executive summary  

SciTech Connect (OSTI)

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

Not Available

1980-09-30T23:59:59.000Z

88

Hanford Low Activity Waste (LAW) Fluidized Bed Steam Reformer (FBSR) Na-Al-Si (NAS) Waste Form Qualification  

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

Hanford Low Activity Waste (LAW) Fluidized Bed Steam Hanford Low Activity Waste (LAW) Fluidized Bed Steam Reformer (FBSR) Na-Al-Si (NAS) Waste Form Qualification C.M. Jantzen and E.M. Pierce November 18, 2010 2 Participating Organizations 3 Incentive and Objectives FBSR sodium-aluminosilicate (NAS) waste form has been identified as a promising supplemental treatment technology for Hanford LAW Objectives: Reduce the risk associated with implementing the FBSR NAS waste form as a supplemental treatment technology for Hanford LAW Conduct test with actual tank wastes Use the best science to fill key data gaps Linking previous and new results together 4 Outline FBSR NAS waste form processing scales FBSR NAS waste form data/key assumptions FBSR NAS key data gaps FBSR NAS testing program 5 FBSR NAS Waste Form Processing

89

West Valley Demonstration Project Waste Management Final Environmental Impact Statement  

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

WVDP Waste Management EIS WVDP Waste Management EIS S-3 Figure S-1. Location of the West Valley Demonstration Project Not to scale Final WVDP Waste Management EIS S-23 Table S-2. Summary of Normal Operational Impacts at West Valley Impact Area Unit of Measure No Action Alternative Alternative A - Preferred Alternative B Human Health Impacts a Public Impacts from Ongoing Operations MEI LCF 3.7 Ă— 10 -7 3.7 Ă— 10 -7 3.7 Ă— 10 -7 Population LCF 1.5 Ă— 10 -3 1.5 Ă— 10 -3 1.5 Ă— 10 -3 Worker Impacts Involved worker MEI LCF 3.4 Ă— 10 -4 1.3 Ă— 10 -3 1.3 Ă— 10 -3 Noninvolved worker MEI LCF 3.0 Ă— 10 -4 3.0 Ă— 10 -4 3.0 Ă— 10 -4 Involved worker population LCF 2.1 Ă— 10 -3 0.031 0.031 Noninvolved worker population LCF 0.075 0.075 0.075 Total worker population LCF 0.077 0.11 0.11

90

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

SciTech Connect (OSTI)

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

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

2010-11-01T23:59:59.000Z

91

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

SciTech Connect (OSTI)

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

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

2013-06-01T23:59:59.000Z

92

RSP WASTE UNIVERSITY OF HAWAII RADIOACTIVE WASTE PICKUP REQUEST FORM Revision 06/07 (WASTE WHICH CONTAINS RADIOISOTOPES BUT NO HAZARDOUS CHEMICALS)  

E-Print Network [OSTI]

RSP WASTE UNIVERSITY OF HAWAII RADIOACTIVE WASTE PICKUP REQUEST FORM Revision 06/07 (WASTE WHICH CONTAINS RADIOISOTOPES BUT NO HAZARDOUS CHEMICALS) INSTRUCTIONS : 1. *NO ISOTOPES MAY BE MIXED IN THE WASTE BOX! One type of isotope per waste box - Except C-14 AND H-3 WHICH MAY BE DISPOSED OF TOGETHER. 2

Browder, Tom

93

E-Print Network 3.0 - acid waste forms Sample Search Results  

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

wastes in the form of gases and ash, often creating entirely new hazards, like dioxins and furans... discussion of waste incineration. Today we know: PCDDF are...

94

Final Tank Closure and Waste Management Environmental Impact...  

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

and treat the waste remaining in 177 underground storage tanks; store the high-level radioactive waste (HLW); dispose of the low-activity waste (LAW) at the Hanford Site...

95

Final Hanford Site Transuranic (TRU) Waste Characterization QA Project Plan  

SciTech Connect (OSTI)

The Transuranic Waste Characterization Quality Assurance Program Plan required each US Department of Energy (DOE) site that characterizes transuranic waste to be sent the Waste Isolation Pilot Plan that addresses applicable requirements specified in the QAPP.

GREAGER, T.M.

1999-09-09T23:59:59.000Z

96

FINAL REPORT WASTE PACKAGE MATERIALS PERFORMANCE PEER REVIEW PANEL  

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

REPORT REPORT WASTE PACKAGE MATERIALS PERFORMANCE PEER REVIEW PANEL FEBRUARY 28, 2002 This document is not an official copy and is for informational purposes only. Signature Page Preface Executive Summary TABLE OF CONTENTS 1. INTRODUCTION 1.1 Organization of the Peer Review 1.2 Objectives of the Review 1.3 Content of the Final Report 2. MAIN FINDINGS 2.1 Perspective 2.2 Overall Findings 2.3 Corrosion Degradation Modes 2.4 Higher or Lower Temperature Operating Modes 2.5 Long-Term Uniform Corrosion of Passive Metal 2.6 Alloy Specification and Comparison 2.7 Technical Issues to be Resolved 2.8 Organizational-Managerial Issues 3. SUMMARY OF DEGRADATION MODES AND CONTRIBUTING FACTORS 3.1 Introduction 3.2 Repository Conditions: Overview of Time, Temperature, Environment

97

AISI waste oxide recycling program. Final technical report  

SciTech Connect (OSTI)

In March 1995 AISI completed a five-year, $60 million collaborative development program on Direct Steelmaking cost-shared by DOE under the Metals Initiative. This program defined an energy-efficient and environmentally-friendly technology to produce hot metal for steelmaking directly from coal and iron ore pellets without incurring the high capital costs and environmental problems associated with traditional coke oven and blast furnace technology. As it becomes necessary to replace present capacity, this new technology will be favored because of reduced capital costs, higher energy efficiency, and lower operating costs. In April 1994, having failed to move forward with a demonstration plant for direct ironmaking, despite substantial efforts by both Stelco and Geneva Steel, an alternative opportunity was sought to commercialize this new technology without waiting until existing ironmaking capacity needed to be replaced. Recycling and resource recovery of steel plant waste oxides was considered an attractive possibility. This led to approval of a ten-month, $8.3 million joint program with DOE on recycling steel plant waste oxides utilizing this new smelting technology. This highly successful trial program was completed in December 1994. The results of the pilot plant work and a feasibility study for a recycling demonstration plant are presented in this final technical report.

Aukrust, E.; Downing, K.B.; Sarma, B.

1995-08-01T23:59:59.000Z

98

Final Hanford Site Transuranic (TRU) Waste Characterization QA Project Plan  

SciTech Connect (OSTI)

The Quality Assurance Project Plan (QAPjP) has been prepared for waste characterization activities to be conducted by the Transuranic (TRU) Project at the Hanford Site to meet requirements set forth in the Waste Isolation Pilot Plan (WIPP) Hazardous Waste Facility Permit, 4890139088-TSDF, Attachment B, including Attachments B1 through B6 (WAP) (DOE, 1999a). The QAPjP describes the waste characterization requirements and includes test methods, details of planned waste sampling and analysis, and a description of the waste characterization and verification process. In addition, the QAPjP includes a description of the quality assurance/quality control (QA/QC) requirements for the waste characterization program. Before TRU waste is shipped to the WIPP site by the TRU Project, all applicable requirements of the QAPjP shall be implemented. Additional requirements necessary for transportation to waste disposal at WIPP can be found in the ''Quality Assurance Program Document'' (DOE 1999b) and HNF-2600, ''Hanford Site Transuranic Waste Certification Plan.'' TRU mixed waste contains both TRU radioactive and hazardous components, as defined in the WLPP-WAP. The waste is designated and separately packaged as either contact-handled (CH) or remote-handled (RH), based on the radiological dose rate at the surface of the waste container. RH TRU wastes are not currently shipped to the WIPP facility.

GREAGER, T.M.

2000-12-06T23:59:59.000Z

99

Transuranic contaminated waste form characterization and data base  

SciTech Connect (OSTI)

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

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

1980-07-01T23:59:59.000Z

100

Basic Research for Evaluating Nuclear Waste Form Performance  

Science Journals Connector (OSTI)

Technical Paper / Argonne National Laboratory Specialists’ Workshop on Basic Research Needs for Nuclear Waste Management / Radioactive Waste

Don J. Bradley

Note: This page contains sample records for the topic "final waste form" 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

Final Hanford Site Transuranic (TRU) Waste Characterization QA Project Plan  

SciTech Connect (OSTI)

The Transuranic Waste Characterization Quality Assurance Program Plan required each U.S. Department of Energy (DOE) site that characterizes transuranic waste to be sent the Waste Isolation Pilot Plan that addresses applicable requirements specified in the quality assurance project plan (QAPP).

GREAGER, T.M.

1999-12-14T23:59:59.000Z

102

Crystalline ceramics: Waste forms for the disposal of weapons plutonium  

SciTech Connect (OSTI)

At present, there are three seriously considered options for the disposition of excess weapons plutonium: (i) incorporation, partial burn-up and direct disposal of MOX-fuel; (ii) vitrification with defense waste and disposal as glass ``logs``; (iii) deep borehole disposal (National Academy of Sciences Report, 1994). The first two options provide a safeguard due to the high activity of fission products in the irradiated fuel and the defense waste. The latter option has only been examined in a preliminary manner, and the exact form of the plutonium has not been identified. In this paper, we review the potential for the immobilization of plutonium in highly durable crystalline ceramics apatite, pyrochlore, monazite and zircon. Based on available data, we propose zircon as the preferred crystalline ceramic for the permanent disposition of excess weapons plutonium.

Ewing, R.C.; Lutze, W. [New Mexico Univ., Albuquerque, NM (United States); Weber, W.J. [Pacific Northwest Lab., Richland, WA (United States)

1995-05-01T23:59:59.000Z

103

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

104

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

SciTech Connect (OSTI)

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

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

2013-01-09T23:59:59.000Z

105

STABILIZING GLASS BONDED WASTE FORMS CONTAINING FISSION PRODUCTS SEPARATED FROM SPENT NUCLEAR FUEL  

SciTech Connect (OSTI)

A model has been developed to represent the stresses developed when a molten, glass-bonded brittle cylinder (used to store nuclear material) is cooled from high temperature to working temperature. Large diameter solid cylinders are formed by heating glass or glass-bonded mixtures (mixed with nuclear waste) to high temperature (915°C). These cylinders must be cooled as the final step in preparing them for storage. Fast cooling time is desirable for production; however, if cooling is too fast, the cylinder can crack into many pieces. To demonstrate the capability of the model, cooling rate cracking data were obtained on small diameter (7.8 cm diameter) glass-only cylinders. The model and experimental data were combined to determine the critical cooling rate which separates the non-cracking stable glass region from the cracked, non-stable glass regime. Although the data have been obtained so far only on small glass-only cylinders, the data and model were used to extrapolate the critical-cooling rates for large diameter ceramic waste form (CWF) cylinders. The extrapolation estimates long term cooling requirements. While a 52-cm diameter cylinder (EBR-II-waste size) can be cooled to 100°C in 70 hours without cracking, a 181.5-cm diameter cylinder (LWR waste size) requires 35 days to cool to 100°C. These cooling times are long enough that verification of these estimates are required so additional experiments are planned on both glass only and CWF material.

Kenneth J. Bateman; Charles W. Solbrig

2008-07-01T23:59:59.000Z

106

Round Robin Testing of the Ceramic Waste Form (CWF)  

SciTech Connect (OSTI)

The Savannah River Technology Center (SRTC) has participated in a round robin testing program, which was conducted under the auspices of the Department of Energy's Tanks Focus Area (TFA) for Immobilization. The round robin, lead by Argonne National Laboratory (ANL), focused on leach testing data of the Ceramic Waste Form (CWF) using the Product Consistency Test (PCT) (ASTM C 1285) and the ANL developed Rapid Water Soluble (RWS) procedure. The CWF is a heterogeneous material comprised of about 70 percent sodalite, 25 percent borosilicate glass binder, 3 percent halite, and 2 percent mixed rare earth and actinide oxides, by mass.

Herman, C.C.

2001-10-02T23:59:59.000Z

107

Naturally occurring crystalline phases: analogues for radioactive waste forms  

SciTech Connect (OSTI)

Naturally occurring mineral analogues to crystalline phases that are constituents of crystalline radioactive waste forms provide a basis for comparison by which the long-term stability of these phases may be estimated. The crystal structures and the crystal chemistry of the following natural analogues are presented: baddeleyite, hematite, nepheline; pollucite, scheelite;sodalite, spinel, apatite, monazite, uraninite, hollandite-priderite, perovskite, and zirconolite. For each phase in geochemistry, occurrence, alteration and radiation effects are described. A selected bibliography for each phase is included.

Haaker, R.F.; Ewing, R.C.

1981-01-01T23:59:59.000Z

108

RSP-MW UNIVERSITY OF HAWAII RADIOACTIVE MIXED WASTE PICKUP REQUEST FORM Revision, 4/04 (WASTE CONTAINING BOTH RADIOISOTOPES AND HAZARDOUS CHEMICALS)  

E-Print Network [OSTI]

RSP-MW UNIVERSITY OF HAWAII RADIOACTIVE MIXED WASTE PICKUP REQUEST FORM Revision, 4/04 (WASTE AND UNDERSTAND ALL CONDITIONS ON THIS FORM. GENERATOR CERTIFICATION: I certify the above waste contains

Browder, Tom

109

Contaminant Release from Residual Waste in Closed Single-Shell Tanks and Other Waste Forms Associated with the Tanks  

SciTech Connect (OSTI)

This chapter describes the release of contaminants from the various waste forms that are anticipated to be associated with closure of the single-shell tanks. These waste forms include residual sludge or saltcake that will remain in the tanks after waste retrieval. Other waste forms include engineered glass and cementitious materials as well as contaminated soil impacted by previous tank leaks. This chapter also describes laboratory testing to quantify contaminant release and how the release data are used in performance/risk assessments for the tank waste management units and the onsite waste disposal facilities. The chapter ends with a discussion of the surprises and lessons learned to date from the testing of waste materials and the development of contaminant release models.

Deutsch, William J.

2008-01-17T23:59:59.000Z

110

Standard test method for static leaching of monolithic waste forms for disposal of radioactive waste  

E-Print Network [OSTI]

1.1 This test method provides a measure of the chemical durability of a simulated or radioactive monolithic waste form, such as a glass, ceramic, cement (grout), or cermet, in a test solution at temperatures radioactive waste forms in various leachants under the specific conditions of the test based on analysis of the test solution. Data from this test are used to calculate normalized elemental mass loss values from specimens exposed to aqueous solutions at temperatures <100°C. 1.3 The test is conducted under static conditions in a constant solution volume and at a constant temperature. The reactivity of the test specimen is determined from the amounts of components released and accumulated in the solution over the test duration. A wide range of test conditions can be used to study material behavior, includin...

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

111

FINAL MEETING SUMMARY HANFORD ADVISORY BOARD TANK WASTE COMMITTEE  

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

or opinions given. Examination of this document cannot equal or replace attendance and public participation. Opening Dirk Dunning, Tank Waste Committee (TWC) chair, welcomed the...

112

Idaho High-Level Waste & Facilities Disposition, Final Environmental...  

Office of Environmental Management (EM)

from INTEC closure activities stored in the Tank Farm, solids in the bottom of the tanks, and trace contamination from first cycle reprocessing extraction waste. SBW contains...

113

Integrated testing of the SRL-165 glass waste form  

SciTech Connect (OSTI)

Integrated testing of the important components of a glass waste form waste package has been performed in order to gain a better understanding of the processes of radionuclide release and transport in the near field environment. Based upon an interpretation of the depth of penetration of hydrogen in reacted SRL-165 glass we have modeled the radionuclide release from the glass as a combined process of (1) the diffusive exchange of alkalis and boron in the glass for hydrogen species in the solution (D = 10{sup -16} cm{sup 2}/s) and (2) surface dissolution. Surface dissolution controls the release of components not exchanged by diffusion and takes place at a rate of 1.5 to 3.0 {mu}m/yr. Subsequent to release the radionuclides may remain in the leach solution, diffuse into the tuff, or precipitate as secondary phases. Precipitation is particularly important for plutonium and americium. Diffusive transport of radionuclides through the tuff takes place at an extremely slow rate, D = 10{sup -16} cm{sup 2}/s. As such, the mass of radionuclides incorporated in the tuff by diffusion during the tests is inconsequential relative to that in the leach solution (with the exception of plutonium) and can be ignored in mass balance calculations. Mass balance calculations based upon the release of radionuclides by surface dissolution of the glass waste form are in good agreement with observed solution chemistry when allowances are made for a pulse of dissolution early in the tests. This pulse may be due to either the rapid dissolution of high-energy surface features early in the integrated tests, or an initially high surface dissolution rate that decreases with time as silica saturation is approached, or a combination of the two.

Phinney, D.L.; Ryerson, F.J.; Oversby, V.M.; Lanford, W.A.; Aines, R.D.; Bates, J.K.

1986-12-01T23:59:59.000Z

114

Waste Isolation Pilot Plant Electronic FOIA Request Form  

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

Request (FOIA) Request (FOIA) Waste Isolation Pilot Plant Electronic FOIA Request Form To make an Electronic FOIA request, please provide the information below. Failure to enter accurate and complete information may render your FOIA request impossible to fulfill. Requests submitted under the Privacy Act must be signed and, therefore, cannot be submitted on this form. Name: Organization: Address: Phone: FAX: Email: Reasonable Describe Records Describe the specific record(s) you seek with sufficient detail that a knowledgeable official of the activity may locate the record with a reasonable amount of effort. Such detail should include: dates, titles, file designations, and offices to be searched. Since most DOE records are not retained permanently, the more information that

115

Stainless steel-zirconium waste forms from the treatment of spent nuclear fuel  

Science Journals Connector (OSTI)

Stainless steel-zirconium waste-form alloys have been developed for the disposal of metallic wastes recovered from spent nuclear fuel using the electrometallurgical process developed by Argonne National Laborator...

S. M. McDeavitt; D. P. Abraham; J. Y. Park; D. D. Keiser

1997-07-01T23:59:59.000Z

116

Final Hazard Categorization for the Remediation of the 116-C-3 Chemical Waste Tanks  

SciTech Connect (OSTI)

This final hazard categorization (FHC) document examines the hazards, identifies appropriate controls to manage the hazards, and documents the commitments for the 116-C-3 Chemical Waste Tanks Remediation Project. The remediation activities analyzed in this FHC are based on recommended treatment and disposal alternatives described in the Engineering Evaluation for the Remediation to the 116-C-3 Chemical Waste Tanks (BHI 2005e).

T. M. Blakley; W. D. Schofield

2007-09-10T23:59:59.000Z

117

MODELING SOLIDIFICATION-INDUCED STRESSES IN CERAMIC WASTE FORMS CONTAINING NUCLEAR WASTES  

SciTech Connect (OSTI)

The goal of this work is to produce a ceramic waste form (CWF) that permanently occludes radioactive waste. This is accomplished by absorbing radioactive salts into zeolite, mixing with glass frit, heating to a molten state 915 C to form a sodalite glass matrix, and solidifying for long-term storage. Less long term leaching is expected if the solidifying cooling rate doesn’t cause cracking. In addition to thermal stress, this paper proposes that a stress is formed during solidification which is very large for fast cooling rates during solidification and can cause severe cracking. A solidifying glass or ceramic cylinder forms a dome on the cylinder top end. The temperature distribution at the time of solidification causes the stress and the dome. The dome height, “the length deficit,” produces an axial stress when the solid returns to room temperature with the inherent outer region in compression, the inner in tension. Large tensions will cause cracking of the specimen. The temperature deficit, derived by dividing the length deficit by the coefficient of thermal expansion, allows solidification stress theory to be extended to the circumferential stress. This paper derives the solidification stress theory, gives examples, explains how to induce beneficial stresses, and compares theory to experimental data.

Charles W. Solbrig; Kenneth J. Bateman

2010-11-01T23:59:59.000Z

118

Final disposal options for mercury/uranium mixed wastes from the Oak Ridge Reservation  

SciTech Connect (OSTI)

Laboratory testing was completed on chemical stabilization and physical encapsulation methods that are applicable (to comply with federal and state regulations) to the final disposal of both hazardous and mixed hazardous elemental mercury waste that is in either of the following categories: (1) waste generated during decontamination and decommissioning (D and D) activities on mercury-contaminated buildings, such as Building 9201-4 at the Oak Ridge Y-12 Plant, or (2) waste stored and regulated under either the Federal Facilities Compliance Agreement or the Federal Facilities Compliance Act. Methods were used that produced copper-mercury, zinc-mercury, and sulfur-mercury materials at room temperature by dry mixing techniques. Toxicity Characteristic Leaching Procedure (TCLP) results for mercury on batches of both the copper-mercury and the sulfur-mercury amalgams consistently produced leachates with less than the 0.2-mg/L Resource Conservation and Recovery Act (RCRA) regulatory limit for mercury. The results clearly showed that the reaction of mercury with sulfur at room temperature produces black mercuric sulfide, a material that is well suited for land disposal. The results also showed that the copper-mercury and zinc-mercury amalgams had major adverse properties that make them undesirable for land disposal. In particular, they reacted readily in air to form oxides and liberate elemental mercury. Another major finding of this study is that sulfur polymer cement is potentially useful as a physical encapsulating agent for mercuric sulfide. This material provides a barrier in addition to the chemical stabilization that further prevents mercury, in the form of mercuric sulfide, from migrating into the environment.

Gorin, A.H.; Leckey, J.H.; Nulf, L.E.

1994-08-29T23:59:59.000Z

119

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

SciTech Connect (OSTI)

The Savannah River National Laboratory (SRNL) was requested to generate a document for the Washington State Department of Ecology and the U.S. Environmental Protection Agency that would cover the following topics: (1) A description of the mineral structures produced by Fluidized Bed Steam Reforming (FBSR) of Hanford type Low Activity Waste (LAW including LAWR which is LAW melter recycle waste) waste, especially the cage structured minerals and how they are formed. (2) How the cage structured minerals contain some contaminants, while others become part of the mineral structure (Note that all contaminants become part of the mineral structure and this will be described in the subsequent sections of this report). (3) Possible contaminant release mechanisms from the mineral structures. (4) Appropriate analyses to evaluate these release mechanisms. (5) Why the appropriate analyses are comparable to the existing Hanford glass dataset. In order to discuss the mineral structures and how they bond contaminants a brief description of the structures of both mineral (ceramic) and vitreous waste forms will be given to show their similarities. By demonstrating the similarities of mineral and vitreous waste forms on atomic level, the contaminant release mechanisms of the crystalline (mineral) and amorphous (glass) waste forms can be compared. This will then logically lead to the discussion of why many of the analyses used to evaluate vitreous waste forms and glass-ceramics (also known as glass composite materials) are appropriate for determining the release mechanisms of LAW/LAWR mineral waste forms and how the durability data on LAW/LAWR mineral waste forms relate to the durability data for LAW/LAWR glasses. The text will discuss the LAW mineral waste form made by FBSR. The nanoscale mechanism by which the minerals form will be also be described in the text. The appropriate analyses to evaluate contaminant release mechanisms will be discussed, as will the FBSR test results to date and how they compare to testing performed on LAW glasses. Other details about vitreous waste form durability and impacts of REDuction/OXidation (REDOX) on durability are given in Appendix A. Details about the FBSR process, various pilot scale demonstrations, and applications are given in Appendix B. Details describing all the different leach tests that need to be used jointly to determine the leaching mechanisms of a waste form are given in Appendix C. Cautions regarding the way in which the waste form surface area is measured and in the choice of leachant buffers (if used) are given in Appendix D.

Jantzen, C

2008-12-26T23:59:59.000Z

120

Final Environmental Impact Statement Safe Interim Storage Of Hanford Tank Wastes  

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

1995/01eis0212_cl.html[6/27/2011 1:02:59 PM] 1995/01eis0212_cl.html[6/27/2011 1:02:59 PM] Final Environmental Impact Statement Safe Interim Storage Of Hanford Tank Wastes DOE/EIS-0212 VOLUME 1 OF 2 VOLUME 1 FINAL ENVIRONMENTAL IMPACT STATEMENT SAFE INTERIM STORAGE OF HANFORD TANK WASTES Hanford Site Richland, Washington October, 1995 WASHINGTON STATE DEPARTMENT OF ECOLOGY NUCLEAR WASTE PROGRAM LACEY, WASHINGTON 98503 U.S. DEPARTMENT OF ENERGY RICHLAND OPERATIONS OFFICE

Note: This page contains sample records for the topic "final waste form" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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121

Material Recovery and Waste Form Development FY 2014 Accomplishments Report  

SciTech Connect (OSTI)

Develop advanced nuclear fuel cycle separation and waste management technologies that improve current fuel cycle performance and enable a sustainable fuel cycle, with minimal processing, waste generation, and potential for material diversion.

Lori Braase

2014-11-01T23:59:59.000Z

122

DOE high-level waste tank safety program. Final report  

SciTech Connect (OSTI)

The overall objective of the work was to provide LANL with support to the DOE High-Level Waste Tank Safety Program. This effort included direct support to the DOE High-Level Waste Tank Working Groups, development of a database to track all identified safety issues, development of requirements for waste tank modernization, evaluation of external comments regarding safety-related guidance/instruction developed previously, examination of current federal and state regulations associated with DOE Tank farm operations, and performance of a conduct of operations review. All tasks which were assigned under this Task Order were completed. Descriptions of the objectives of each task and effort performed to complete each objective is provided.

NONE

1998-11-01T23:59:59.000Z

123

Annual report on the development and characterization of solidified forms for nuclear wastes, 1979  

SciTech Connect (OSTI)

Development and characterization of solidified nuclear waste forms is a major continuing effort at Pacific Northwest Laboratory. Contributions from seven programs directed at understanding chemical composition, process conditions, and long-term behaviors of various nuclear waste forms are included in this report. The major findings of the report are included in extended figure captions that can be read as brief technical summaries of the research, with additional information included in a traditional narrative format. Waste form development proceeded on crystalline and glass materials for high-level and transuranic (TRU) wastes. Leaching studies emphasized new areas of research aimed at more basic understanding of waste form/aqueous solution interactions. Phase behavior and thermal effects research included studies on crystal phases in defense and TRU waste glasses and on liquid-liquid phase separation in borosilicate waste glasses. Radiation damage effects in crystals and glasses from alpha decay and from transmutation are reported.

Chick, L.A.; McVay, G.L.; Mellinger, G.B.; Roberts, F.P.

1980-12-01T23:59:59.000Z

124

Vitrified municipal waste as a host form for high-level nuclear waste  

Science Journals Connector (OSTI)

Using glass as a safe and long term hosting matrix for hazardous wastes and for the immobilization of heavy metals and nuclear wastes has become an attractive method [3]. The most known glasses used as nuclear waste

N. A. El-Alaily; E. M. Abou-Hussein…

2014-01-01T23:59:59.000Z

125

Steam reforming of low-level mixed waste. Final report  

SciTech Connect (OSTI)

ThermoChem has successfully designed, fabricated and operated a nominal 90 pound per hour Process Development Unit (PDU) on various low-level mixed waste surrogates. The design, construction, and testing of the PDU as well as performance and economic projections for a 300-lb/hr demonstration and commercial system are described. The overall system offers an environmentally safe, non-incinerating, cost-effective, and publicly acceptable method of processing LLMW. The steam-reforming technology was ranked the No. 1 non-incineration technology for destruction of hazardous organic wastes in a study commissioned by the Mixed Waste Focus Area and published in April 1997. The ThermoChem steam-reforming system has been developed over the last 13 years culminating in this successful test campaign on LLMW surrogates. Six surrogates were successfully tested including a 750-hour test on material simulating a PCB- and Uranium-contaminated solid waste found at the Portsmouth Gaseous Diffusion Plant. The test results indicated essentially total (> 99.9999%) destruction of RCRA and TSCA hazardous halogenated organics, significant levels of volume reduction (> 400 to 1), and retention of radionuclides in the volume-reduced solids. Economic evaluations have shown the steam-reforming system to be very cost competitive with more conventional and other emerging technologies.

NONE

1998-06-01T23:59:59.000Z

126

Speciation of heavy metals in cement-stabilized waste forms: A micro-spectroscopic study  

E-Print Network [OSTI]

Assuring safe disposal and long-term storage of haz- ardous and radioactive wastes represents a primary en- vironmental task of industrial societies. The long-term disposal of the hazardous wastes is associatedSpeciation of heavy metals in cement-stabilized waste forms: A micro-spectroscopic study M. Vespa

127

Taiwan industrial cooperation program technology transfer for low-level radioactive waste final disposal - phase I.  

SciTech Connect (OSTI)

Sandia National Laboratories and the Institute of Nuclear Energy Research, Taiwan have collaborated in a technology transfer program related to low-level radioactive waste (LLW) disposal in Taiwan. Phase I of this program included regulatory analysis of LLW final disposal, development of LLW disposal performance assessment capabilities, and preliminary performance assessments of two potential disposal sites. Performance objectives were based on regulations in Taiwan and comparisons to those in the United States. Probabilistic performance assessment models were constructed based on limited site data using software including GoldSim, BLT-MS, FEHM, and HELP. These software codes provided the probabilistic framework, container degradation, waste-form leaching, groundwater flow, radionuclide transport, and cover infiltration simulation capabilities in the performance assessment. Preliminary performance assessment analyses were conducted for a near-surface disposal system and a mined cavern disposal system at two representative sites in Taiwan. Results of example calculations indicate peak simulated concentrations to a receptor within a few hundred years of LLW disposal, primarily from highly soluble, non-sorbing radionuclides.

Knowlton, Robert G.; Cochran, John Russell; Arnold, Bill Walter; Jow, Hong-Nian; Mattie, Patrick D.; Schelling, Frank Joseph Jr. (; .)

2007-01-01T23:59:59.000Z

128

Environmental assessment for transuranic waste work-off plan, Los Alamos National Laboratory. Rough draft: Final report  

SciTech Connect (OSTI)

The Los Alamos National Laboratory (LANL) generates transuranic (TRU) waste in a variety of programs related to national defense. TRU waste is a specific class of radioactive waste requiring permanent isolation. Most defense-related TRU waste will be permanently disposed of in the Waste Isolation Pilot Plant (WIPP). WIPP is a deep geologic repository located in southeastern New Mexico and is now in the testing phase of development. All waste received by Wipp must conform with established Waste Acceptance Criteria (WAC). The purpose of the proposed action is to retrieve stored TRU waste and prepare the waste for shipment to and disposal WIPP. Stored TRU waste LANL is represented by four waste forms. The facilities necessary for work-off activities are tailored to the treatment and preparation of these four waste forms. Preparation activities for newly generated TRU waste are also covered by this action.

Not Available

1990-10-26T23:59:59.000Z

129

Radioactive Material Declaration Form Exhibit to the Radioactive Waste Manual (RWM)  

E-Print Network [OSTI]

Radioactive Material Declaration Form Exhibit to the Radioactive Waste Manual (RWM) 12/5/2013 (form Declaration Form Exhibit to the Radioactive Waste Manual (RWM) 12/5/2013 (form date) SLAC-I-760-2A08Z-001 (RWM date) SLAC-I-760-2A08Z-001 (RWM number) Page 1 of 2 RADIOACTIVE MATERIAL DECLARATION FORM For RP use

Wechsler, Risa H.

130

Waste Isolation Pilot Plant design validation: Final report, Appendices  

SciTech Connect (OSTI)

This volume is comprised of the following appendices: DOE stipulated agreement with State of New Mexico (partial); geologic correlations; mathematical simulation of underground in situ behavior; C and SH shaft geologic logs and maps; waste shaft geologic logs and maps; exhaust shaft geologic log; test rooms geologic maps and sections; drift cross sections; facility level geologic core hole logs; geomechanical instrumentation data plots; and analytical data plots.

Not Available

1986-10-01T23:59:59.000Z

131

Macroencapsulation of mixed waste debris at the Hanford Nuclear Reservation -- Final project report by AST Environmental Services, LLC  

SciTech Connect (OSTI)

This report summarizes the results of a full-scale demonstration of a high density polyethylene (HDPE) package, manufactured by Arrow Construction, Inc. of Montgomery, Alabama. The HDPE package, called ARROW-PAK, was designed and patented by Arrow as both a method to macroencapsulation of radioactively contaminated lead and as an improved form of waste package for treatment and interim and final storage and/or disposal of drums of mixed waste. Mixed waste is waste that is radioactive, and meets the criteria established by the United States Environmental Protection Agency (US EPA) for a hazardous material. Results from previous testing conducted for the Department of Energy (DOE) at the Idaho National Engineering Laboratory in 1994 found that the ARROW-PAK fabrication process produces an HDPE package that passes all helium leak tests and drop tests, and is fabricated with materials impervious to the types of environmental factors encountered during the lifetime of the ARROW-PAK, estimated to be from 100 to 300 years. Arrow Construction, Inc. has successfully completed full-scale demonstration of its ARROW-PAK mixed waste macroencapsulation treatment unit at the DOE Hanford Site. This testing was conducted in accordance with Radiological Work Permit No. T-860, applicable project plans and procedures, and in close consultation with Waste Management Federal Services of Hanford, Inc.`s project management, health and safety, and quality assurance representatives. The ARROW-PAK field demonstration successfully treated 880 drums of mixed waste debris feedstock which were compacted and placed in 149 70-gallon overpack drums prior to macroencapsulation in accordance with the US EPA Alternate Debris Treatment Standards, 40 CFR 268.45. Based on all of the results, the ARROW-PAK process provides an effective treatment, storage and/or disposal option that compares favorably with current mixed waste management practices.

Baker, T.L.

1998-02-25T23:59:59.000Z

132

Separation of technetium from nuclear waste stream simulants. Final report  

SciTech Connect (OSTI)

The author studied liquid anion exchangers, such as Aliquat-336 nitrate, various pyridinium nitrates, and related salts, so that they may be applied toward a specific process for extracting (partitioning) and recovering {sup 99}TcO{sub 4}{sup {minus}} from nuclear waste streams. Many of the waste streams are caustic and contain a variety of other ions. For this reason, the author studied waste stream simulants that are caustic and contain appropriate concentrations of selected, relevant ions. Methods of measuring the performance of the exchangers and extractant systems included contact experiments. Batch contact experiments were used to determine the forward and reverse extraction parameters as a function of temperature, contact time, phase ratio, concentration, solvent (diluent), and other physical properties. They were also used for stability and competition studies. Specifically, the author investigated the solvent extraction behavior of salts of perrhenate (ReO{sub 4}{sup {minus}}), a stable (non-radioactive) chemical surrogate for {sup 99}TcO{sub 4}{sup {minus}}. Results are discussed for alternate organic solvents; metalloporphyrins, ferrocenes, and N-cetyl pyridium nitrate as alternate extractant salts; electroactive polymers; and recovery of ReO{sub 4}{sup {minus}} and TcO{sub 4}{sup {minus}}.

Strauss, S.H. [Colorado State Univ., Fort Collins, CO (United States). Dept. of Chemistry

1995-09-11T23:59:59.000Z

133

Defense High-Level Waste Leaching Mechanisms Program. Final report  

SciTech Connect (OSTI)

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

Mendel, J.E. (compiler)

1984-08-01T23:59:59.000Z

134

Final Hanford Site Solid (Radioactive and Hazardous) Waste Program Environmental Impact Statement Richland, Washington  

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

HSW HSW EIS January 2004 1.6 Figure 1.2. States with Radioactive Waste Disposal Activities Final HSW EIS January 2004 1.12 Figure 1.3. Relationship of the HSW EIS to Other Hanford Cleanup Operations, Material Management Activities, and Key Environmental Reviews 2.17 Final HSW EIS January 2004 Figure 2.6. Waste Receiving and Processing Facility Figure 2.7. X-Ray Image of Transuranic Waste Drum Contents M0212-0286.11 HSW EIS 12-10-02 M0212-0286.12 HSW EIS 12-10-02 2.17 Final HSW EIS January 2004 Figure 2.6. Waste Receiving and Processing Facility Figure 2.7. X-Ray Image of Transuranic Waste Drum Contents M0212-0286.11 HSW EIS 12-10-02 M0212-0286.12 HSW EIS 12-10-02 Final HSW EIS January 2004 2.34 Figure 2.18. Typical Liner System Final HSW EIS January 2004 2.36

135

Secondary Waste Form Development and Optimization—Cast Stone  

SciTech Connect (OSTI)

Washington River Protection Services is considering the design and construction of a Solidification Treatment Unit (STU) for the Effluent Treatment Facility (ETF) at Hanford. The ETF is a Resource Conservation and Recovery Act-permitted, multi-waste, treatment and storage unit and can accept dangerous, low-level, and mixed wastewaters for treatment. The STU needs to be operational by 2018 to receive secondary liquid wastes generated during operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The STU to ETF will provide the additional capacity needed for ETF to process the increased volume of secondary wastes expected to be produced by WTP.

Sundaram, S. K.; Parker, Kent E.; Valenta, Michelle M.; Pitman, Stan G.; Chun, Jaehun; Chung, Chul-Woo; Kimura, Marcia L.; Burns, Carolyn A.; Um, Wooyong; Westsik, Joseph H.

2011-07-14T23:59:59.000Z

136

An experimental survey of the factors that affect leaching from low-level radioactive waste forms  

SciTech Connect (OSTI)

This report represents the results of an experimental survey of the factors that affect leaching from several types of solidified low-level radioactive waste forms. The goal of these investigations was to determine those factors that accelerate leaching without changing its mechanism(s). Typically, although not in every case,the accelerating factors include: increased temperature, increased waste loading (i.e., increased waste to binder ratio), and decreased size (i.e., decreased waste form volume to surface area ratio). Additional factors that were studied were: increased leachant volume to waste form surface area ratio, pH, leachant composition (groundwaters, natural and synthetic chelating agents), leachant flow rate or replacement frequency and waste form porosity and surface condition. Other potential factors, including the radiation environment and pressure, were omitted based on a survey of the literature. 82 refs., 236 figs., 13 tabs.

Dougherty, D.R.; Pietrzak, R.F.; Fuhrmann, M.; Colombo, P.

1988-09-01T23:59:59.000Z

137

A Method to Evaluate Additional Waste Forms to Optimize Performance of the HLW Repository  

SciTech Connect (OSTI)

The DOE high-level waste (HLW) disposal system is based on decisions made in the 1970s. The de facto Yucca Mountain WAC for HLW, contained in the Waste Acceptance System Requirements Document (WASRD), and the DOE-EM Waste Acceptance Product Specification for Vitrified High Level Waste Forms (WAPS) tentatively describes waste forms to be interred in the repository, and limits them to borosilicate glass (BSG). It is known that many developed waste forms are as durable as or better than environmental assessment or “EA”-glass. Among them are the salt-ceramic and metallic waste forms developed at ANL-W. Also, iron phosphate glasses developed at University of Missouri show promise in stabilizing the most refractory materials in Hanford HLW. However, for any of this science to contribute, the current Total System Performance Assessment model must be able to evaluate the additional waste form to determine potential impacts on repository performance. The results can then support the technical bases required in the repository license application. A methodology is proposed to use existing analysis models to evaluate potential additional waste forms for disposal without gathering costly material specific degradation data. The concept is to analyze the potential impacts of waste form chemical makeup on repository performance assuming instantaneous waste matrix dissolution. This assumption obviates the need for material specific degradation models and is based on the relatively modest fractional contribution DOE HLW makes to the repository radionuclide and hazardous metals inventory. The existing analysis models, with appropriate data modifications, are used to evaluate geochemical interactions and material transport through the repository. This methodology would support early screening of proposed waste forms through simplified evaluation of disposal performance, and would provide preliminary guidance for repository license amendment in the future.

D. Gombert; L. Lauerhass

2006-02-01T23:59:59.000Z

138

Immobilization of fission products in phosphate ceramic waste forms  

SciTech Connect (OSTI)

Argonne National Laboratory (ANL) is developing chemically bonded phosphate ceramics (CBPCs) to treat low-level mixed wastes, particularly those containing volatiles and pyrophorics that cannot be treated by conventional thermal processes. This work was begun under ANL`s Laboratory Directed Research and Development funds, followed by further development with support from EM-50`s Mixed Waste Focus Area.

Singh, D.; Wagh, A.

1997-09-01T23:59:59.000Z

139

Final Reclamation Report: Basalt Waste Isolation Project exploratory shaft site  

SciTech Connect (OSTI)

The restoration of areas disturbed by activities of the Basalt Waste Isolation Project (BWIP) constitutes a unique operation at the US Department of Energy's (DOE) Hanford Site, both from the standpoint of restoration objectives and the time frame for accomplishing these objectives. The BWIP reclamation program comprises three separate projects: borehole reclamation, Near Surface Test Facility (NSTF) reclamation, and Exploratory Shaft Facility (ESF) reclamation. The main focus of this report is on determining the success of the revegetation effort 1 year after work was completed. This report also provides a brief overview of the ESF reclamation program. 21 refs., 7 figs., 14 tabs.

Brandt, C.A.; Rickard, W.H. Jr.

1990-06-01T23:59:59.000Z

140

Rural electrification: Waste biomass Russian northern territories. Final report  

SciTech Connect (OSTI)

The primary objective of this pre-feasibility evaluation is to examine the economic and technical feasibility of replacing distillate fuel with local waste biomass in the village of Verkhni-Ozerski, Arkhangelsk Region, Russia. This village is evaluated as a pilot location representing the off-grid villages in the Russian Northern Territories. The U.S. Department of Energy (DOE) has agreed to provide technical assistance to the Ministry of Fuel and Energy (MFE). MFE has identified the Northern Territories as a priority area requiring NREL`s assistance. The program initially affects about 900 off-grid villages. Biomass and wind energy, and to a lesser extent small hydro (depending on resource availability) are expected to play the dominant role in the program, Geothermal energy may also have a role in the Russian Far East. The Arkhangelsk, Kariela, and Krasnoyarsk Regions, all in the Russian Northern Territories, have abundant forest resources and forest products industries, making them strong candidates for implementation of small-scale waste biomass-to-energy projects. The 900 or so villages included in the renewable energy program span nine administrative regions and autonomous republics. The regional authorities in the Northern Territories proposed these villages to MFE for consideration in the renewable energy program according to the following selection criteria: (a) Remote off-grid location, (b) high cost of transporting fuel, old age of existing power generation equipment, and (d) preliminary determination as to availability of alternative energy resources. Inclusion of indigenous minorities in the program was also heavily emphasized. The prefeasibility study demonstrates that the project merits continuation and a full feasibility analysis. The demonstrated rate of return and net positive cash flow, the willingness of Onegales and local/regional authorities to cooperate, and the immense social benefits are all good reasons to continue the project.

Adamian, S. [ECOTRADE, Inc., Glendale, CA (United States)

1998-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "final waste form" 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

Development of long-term performance models for radioactive waste forms  

SciTech Connect (OSTI)

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

Bacon, Diana H.; Pierce, Eric M.

2011-03-22T23:59:59.000Z

142

Final closure cover for a Hanford radioactive mixed waste disposal facility  

SciTech Connect (OSTI)

This study provides a preliminary design for a RCRA mixed waste landfill final closure cover. The cover design was developed by a senior class design team from Seattle University. The design incorporates a layered design of indigenous soils and geosynthetics in a layered system to meet final closure cover requirements for a landfill as imposed by the Washington Administrative Code WAC-173-303 implementation of the Resource Conservation and Recovery Act.

Johnson, K.D.

1996-02-06T23:59:59.000Z

143

State-of-the-art review of materials properties of nuclear waste forms.  

SciTech Connect (OSTI)

The Materials Characterization Center (MCC) was established at the Pacific Northwest Laboratory to assemble a standardized nuclear waste materials data base for use in research, systems and facility design, safety analyses, and waste management decisions. This centralized data base will be provided through the means of a Nuclear Waste Materials Handbook. The first issue of the Handbook will be published in the fall of 1981 in looseleaf format so that it can be updated as additional information becomes available. To ensure utmost reliability, all materials data appearing in the Handbook will be obtained by standard procedures defined in the Handbook and approved by an independent Materials Review Board (MRB) comprised of materials experts from Department of Energy laboratories and from universities and industry. In the interim before publication of the Handbook there is need for a report summarizing the existing materials data on nuclear waste forms. This review summarizes materials property data for the nuclear waste forms that are being developed for immobilization of high-level radioactive waste. It is intended to be a good representation of the knowledge concerning the properties of HLW forms as of March 1981. The table of contents lists the following topics: introduction which covers waste-form categories, and important waste-form materials properties; physical properties; mechanical properties; chemical durability; vaporization; radiation effects; and thermal phase stability.

Mendel, J. E.; Nelson, R. D.; Turcotte, R. P.; Gray, W. J.; Merz, M. D.; Roberts, F. P.; Weber, W. J.; Westsik, Jr., J. H.; Clark, D. E.

1981-04-01T23:59:59.000Z

144

DOE Issues Final Request for Proposal for Oak Ridge Transuranic Waste Processing Center Services  

Broader source: Energy.gov [DOE]

Cincinnati -- The U.S. Department of Energy (DOE) today issued a Final Request for Proposal (RFP), for support services at the Oak Ridge Transuranic Waste Processing Center (TWPC) in Oak Ridge, Tennessee. The total estimated value of the contract is $100 Million - $300 Million.

145

Phase 5 storage (Project W-112) Central Waste Complex operational readiness review, final report  

SciTech Connect (OSTI)

This document is the final report for the RFSH conducted, Contractor Operational Readiness Review (ORR) for the Central Waste Complex (CWC) Project W-112 and Interim Safety Basis implementation. As appendices, all findings, observations, lines of inquiry and the implementation plan are included.

Wight, R.H.

1997-05-30T23:59:59.000Z

146

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

SciTech Connect (OSTI)

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

Veazey, G.W. [Los Alamos National Labs., NM (United States); Ames, R.L. [Rocky Flats Environmental Technology Site, Golden, CO (United States)

1997-03-01T23:59:59.000Z

147

Effects of tuff waste package components on release from 76-68 simulated waste glass: Final report  

SciTech Connect (OSTI)

An experimental matrix has been conducted that will allow evaluation of the effects of waste package constituents on the waste form release behavior in a tuff repository environment. Tuff rock and groundwater were used along with 304L, 316, and 1020M ferrous metals to evaluate release from uranium-doped MCC 76-68 simulated waste glass. One of the major findings was that in the absence of 1020M mild steel, tuff rock powder dominates the system. However, when 1020M mild steel is present, it appears to dominate the system. The rock-dominated system results in suppressed glass-water reaction and leaching while the 1020M-dominated system results in enhanced leaching - but the metal effectively scavenges uranium from solution. The 300-series stainless steels play no significant role in affecting glass leaching characteristics. 6 refs., 28 figs., 5 tabs.

McVay, G.L.; Robinson, G.R.

1984-04-01T23:59:59.000Z

148

Surface effects of cement-based solidified waste forms  

E-Print Network [OSTI]

the proper size aggregate now using 400-Itm and 250-lrm meshed sieves to attain the 250 Itm sand. The radionuclide to be used in the mixture had to be representative of actual simulated low-level radioactive waste. The nuclide also had to have a sufficient... Selection of the counting equipment was based on three criteria: effectiveness of counting the specified radionuclide, the experimenter's knowledge of the system, and the availability of the chosen system. Since the nuclide chosen as the simulated waste...

Pavlonnis, George

2012-06-07T23:59:59.000Z

149

Fissile material disposition program final immobilization form assessment and recommendation  

SciTech Connect (OSTI)

Lawrence Livermore National Laboratory (LLNL), in its role as the lead laboratory for the development of plutonium immobilization technologies for the Department of Energy`s Office of Fissile Materials Disposition (MD), has been requested by MD to recommend an immobilization technology for the disposition of surplus weapons- usable plutonium. The recommendation and supporting documentation was requested to be provided by September 1, 1997. This report addresses the choice between glass and ceramic technologies for immobilizing plutonium using the can-in-canister approach. Its purpose is to provide a comparative evaluation of the two candidate technologies and to recommend a form based on technical considerations.

Cochran, S.G.; Dunlop, W.H.; Edmunds, T.A.; MacLean, L.M.; Gould, T.H. [Westinghouse Savannah River Co., Aiken, SC (United States)

1997-10-03T23:59:59.000Z

150

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

SciTech Connect (OSTI)

Demonstrating that a waste form produced by a given immobilization process is chemically and physically durable as well as compliant with disposal facility acceptance criteria is critical to the success of a waste treatment program, and must be pursued in conjunction with the maturation of the waste processing technology. Testing of waste forms produced using differing scales of processing units and classes of feeds (simulants versus actual waste) is the crux of the waste form qualification process. Testing is typically focused on leachability of constituents of concern (COCs), as well as chemical and physical durability of the waste form. A principal challenge regarding testing immobilized low-activity waste (ILAW) forms is the absence of a standard test suite or set of mandatory parameters against which waste forms may be tested, compared, and qualified for acceptance in existing and proposed nuclear waste disposal sites at Hanford and across the Department of Energy (DOE) complex. A coherent and widely applicable compliance strategy to support characterization and disposal of new waste forms is essential to enhance and accelerate the remediation of DOE tank waste. This paper provides a background summary of important entities, regulations, and considerations for nuclear waste form qualification and disposal. Against this backdrop, this paper describes a strategy for meeting and demonstrating compliance with disposal requirements emphasizing the River Protection Project (RPP) Integrated Disposal Facility (IDF) at the Hanford Site and the fluidized bed steam reforming (FBSR) mineralized low-activity waste (LAW) product stream.

SAMS TL; EDGE JA; SWANBERG DJ; ROBBINS RA

2011-01-13T23:59:59.000Z

151

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

SciTech Connect (OSTI)

This is a milestone document covering the activities to validate theoretical calculations with experimental data for the effect of the decay of 90Sr to 90Zr on materials properties. This was done for a surragate waste form strontium titanate.

Van Ginhoven, Renee M.; Jaffe, John E.; Jiang, Weilin; Strachan, Denis M.

2011-04-01T23:59:59.000Z

152

Final remote-handled waste canister leaves Los Alamos National Laboratory  

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

Remote-handled waste canister leaves LANL Remote-handled waste canister leaves LANL Final remote-handled waste canister leaves Los Alamos National Laboratory The Laboratory began shipping the canisters exactly one month ago and averaged four shipments per week. July 2, 2009 Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma physics and new materials. Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma physics and new materials. Contact

153

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

SciTech Connect (OSTI)

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

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

2013-02-01T23:59:59.000Z

154

Final Environmental Impact Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina  

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

Els-o120 Els-o120 Final Environmental Impact Statement I Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina of Energy 1 COVER SHEET RESPONSIBLE AGENCY: U.S. Department of Energy ACTIVITY: Final Environmental Impact Statement, Waste Management I TC Activities for Groundwater Protection at the Savannah River Plant, Aiken, South Carolina. CONTACT: ABSTRACT: Additional information concerning this Statement can be obtained from: Mr. S. R. Wright Director, Environmental Division U.S. Department of Energy Savannah River Operations Office Post Office Box A Aiken, South Carolina 29802 (803) 725-3957 I TC For general information on the Department of Energy qs EIS process contact: Office of the Assistant Secretary for Environment, Safety, and Health U.S. Department of Energy Attn: Ms. Carol Bergstrom (EH-25) Acting Director, Office of

155

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

SciTech Connect (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

156

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

SciTech Connect (OSTI)

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

Jones, T.L. [New Mexico State Univ., Las Cruces, NM (United States); Serne, R.J. [Pacific Northwest Lab., Richland, WA (United States)

1994-10-01T23:59:59.000Z

157

Microstructural characterization of halite inclusion in a glass-bonded ceramic waste form.  

SciTech Connect (OSTI)

A glass-bonded ceramic waste form is being developed to immobilize radioactively contaminated chloride waste salts generated during the conditioning of spent sodium-bonded nuclear fuel for disposal. The waste salt is first mixed with zeolite A to occlude the salt into cavities in the zeolite structure. The salt-loaded zeolite is then mixed with a borosilicate glass and consolidated by hot isostatic pressing. During this process, the zeolite converts to the mineral sodalite, which retains most of the waste salt, and small amounts of halite are generated. Halite inclusions have been observed within micron- to submicron-sized pores that form within the glass phase in the vicinity of the sodalite/glass interface. These inclusions are important because they may contain small amounts of radionuclide contaminants (eg {sup 135}Cs and {sup 129}I),and may affect the corrosion behavior of the waste form. Optical microscopy, scanning electron microscopy, and transmission electron microscopy were used to characterize the chemical nature and distribution of halite inclusions in the waste form.

Luo, J. S.; Ebert, W. L.

2000-12-14T23:59:59.000Z

158

Epsilon metal waste form for immobilization of noble metals from used nuclear fuel  

Science Journals Connector (OSTI)

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

Jarrod V. Crum; Denis Strachan; Aashish Rohatgi; Mac Zumhoff

2013-01-01T23:59:59.000Z

159

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

SciTech Connect (OSTI)

This report describes the results from laboratory tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate candidate stabilization technologies that have the potential to successfully treat liquid secondary waste stream effluents produced by the Hanford Tank Waste Treatment and Immobilization Plant (WTP). WRPS is considering the design and construction of a Solidification Treatment Unit (STU) for the Effluent Treatment Facility (ETF) at Hanford. The ETF, a multi-waste, treatment-and-storage unit that has been permitted under the Resource Conservation and Recovery Act (RCRA), can accept dangerous, low-level, and mixed wastewaters for treatment. The STU needs to be operational by 2018 to receive secondary liquid waste generated during operation of the WTP. The STU will provide the additional capacity needed for ETF to process the increased volume of secondary waste expected to be produced by WTP. This report on radionuclide retention mechanisms describes the testing and characterization results that improve understanding of radionuclide retention mechanisms, especially for pertechnetate, {sup 99}TcO{sub 4}{sup -} in four different waste forms: Cast Stone, DuraLith alkali aluminosilicate geopolymer, encapsulated fluidized bed steam reforming (FBSR) product, and Ceramicrete phosphate bonded ceramic. These data and results will be used to fill existing data gaps on the candidate technologies to support a decision-making process that will identify a subset of the candidate waste forms that are most promising and should undergo further performance testing.

Um, Wooyong; Valenta, Michelle M.; Chung, Chul-Woo; Yang, Jungseok; Engelhard, Mark H.; Serne, R. Jeffrey; Parker, Kent E.; Wang, Guohui; Cantrell, Kirk J.; Westsik, Joseph H.

2011-09-26T23:59:59.000Z

160

GTS Duratek, phase I Hanford low-level waste melter tests: Final report  

SciTech Connect (OSTI)

A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense waste stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the final report on testing performed by GTS Duratek Inc. in Columbia, Maryland. GTS Duratek (one of the seven vendors selected) was chosen to demonstrate Joule heated melter technology under WHC subcontract number MMI-SVV-384215. The report contains description of the tests, observations, test data and some analysis of the data as it pertains to application of this technology for LLW vitrification. The document also contains summaries of the melter offgas reports issued as separate documents for the 100 kg melter (WHC-SD-WM-VI-028) and for the 1000 kg melter (WHC-SD-WM-VI-029).

Eaton, W.C.

1995-10-26T23:59:59.000Z

Note: This page contains sample records for the topic "final waste form" 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

Summary Report: Glass-Ceramic Waste Forms for Combined Fission Products  

SciTech Connect (OSTI)

Glass-ceramic waste form development began in FY 2010 examining two combined waste stream options: (1) alkaline earth (CS) + lanthanide (Ln), and (2) + transition metal (TM) fission-product waste streams generated by the uranium extraction (UREX+) separations process. Glass-ceramics were successfully developed for both options however; Option 2 was selected over Option 1, at the conclusion of 2010, because Option 2 immobilized all three waste streams with only a minimal decrease in waste loading. During the first year, a series of three glass (Option 2) were fabricated that varied waste loading-WL (42, 45, and 50 mass%) at fixed molar ratios of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali both at 1.75. These glass-ceramics were slow cooled and characterized in terms of phase assemblage and preliminary irradiation stability. This fiscal year, further characterization was performed on the FY 2010 Option 2 glass-ceramics in terms of: static leach testing, phase analysis by transmission electron microscopy (TEM), and irradiation stability (electron and ion). Also, a new series of glass-ceramics were developed for Option 2 that varied the additives: Al{sub 2}O{sub 3} (0-6 mass%), molar ratio of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali (1.75 to 2.25) and waste loading (50, 55, and 60 mass%). Lastly, phase pure powellite and oxyapatite were synthesized for irradiation studies. Results of this fiscal year studies showed compositional flexibility, chemical stability, and radiation stability in the current glass-ceramic system. First, the phase assemblages and microstructure of all of the FY 2010 and 2011 glass-ceramics are very similar once subjected to the slow cool heat treatment. The phases identified in these glass-ceramics were oxyapatite, powellite, cerianite, and ln-borosilicate. This shows that variations in waste loading or additives can be accommodated without drastically changing the phase assemblage of the waste form, thus making the processing and performance characteristics of the waste form more predictable/flexible. However, in the future, the glass phase still needs to be accurately characterized to determine the effects of waste loading and additives on the glass structure. Initial investigations show a borosilicate glass phase rich in silica. Second, the normalized concentrations of elements leached from the waste form during static leach testing were all below 0.6 g/L after 28d at 90 C, by the Product Consistency Test (PCT), method B. These normalized concentrations are on par with durable waste glasses such as the Low-Activity Reference Material (LRM) glass. The release rates for the crystalline phases (oxyapatite and powellite) appear to be lower (more durable) than the glass phase based on the relatively low release rates of Mo, Ca, and Ln found in the crystalline phases compared to Na and B that are mainly observed in the glass phase. However, further static leach testing on individual crystalline phases is needed to confirm this statement. Third, Ion irradiation and In situ TEM observations suggest that these crystalline phases (such as oxyapatite, ln-borosilicate, and powellite) in silicate based glass ceramic waste forms exhibit stability to 1000 years at anticipated doses (2 x 10{sup 10}-2 x 10{sup 11} Gy). This is adequate for the short lived isotopes in the waste, which lead to a maximum cumulative dose of {approx}7 x 10{sup 9} Gy, reached after {approx}100 yrs, beyond which the dose contributions are negligible. The cumulate dose calculations are based on a glass-ceramic at WL = 50 mass%, where the fuel has a burn-up of 51GWd/MTIHM, immobilized after 5 yr decay from reactor discharge.

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

2011-09-23T23:59:59.000Z

162

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

SciTech Connect (OSTI)

An accelerated alpha-decay damage study of a glass-bonded sodalite ceramic waste form has recently been completed. The purpose of this study was to investigate the physical and chemical durability of the waste form after significant exposure to alpha decay. This accelerated alpha-decay study was performed by doping the ceramic waste form with {sup 238}Pu which has a much greater specific activity than {sup 239}Pu that is normally present in the waste form. The alpha-decay dose at the end of the four year study was approximately 1 x 10{sup 18} alpha-decays/gram of material. An equivalent time period for a similar dose of {sup 239}Pu would require approximately 1100 years. After four years of exposure to {sup 238}Pu alpha decay, the investigation observed little change to the physical or chemical durability of the ceramic waste form (CWF). Specifically, the {sup 238}Pu-loaded CWF maintained it's physical integrity, namely that the density remained constant and no cracking or phase de-bonding was observed. The materials chemical durability and phase stability also did not change significantly over the duration of the study. The only significant measured change was an increase of the unit-cell lattice parameters of the plutonium oxide and sodalite phases of the material and an increase in the release of salt components and plutonium of the waste form during leaching tests, but, as mentioned, these did not lead to any overall loss of waste form durability. The principal findings from this study are: (1) {sup 238}Pu-loaded CWF is similar in microstructure and phase composition to referenced waste form. (2) Pu was observed primarily as oxide comprised of aggregates of nano crystals with aggregates ranging in size from submicron to twenty microns in diameter. (3) Pu phases were primarily found in the intergranular glassy regions. (4) PuO phase shows expected unit cell volume expansion due to alpha decay damage of approximately 0.7%, and the sodalite phase unit cell volume has expanded slightly by 0.3% again, presumably due to alpha-decay damage. (5) No bulk sample swelling was observed. (6) No amorphization of sodalite or actinide bearing phases was observed after four years of alpha-decay damage. (7) No microcracks or phase de-bonding were observed in waste form samples aged for four years. (8) In some areas of the {sup 238}Pu doped ceramic waste form material bubbles and voids were found. Bubbles and voids with similar size and density were also found in ceramic waste form samples without actinide. These bubbles and voids are interpreted as pre-existing defects. However, some contribution to these bubbles and voids from helium gas can not be ruled out. (9) Chemical durability of {sup 238}Pu CWF has not changed significantly after four years of alpha-decay exposure except for an increase in the release of salt components and Pu. Still, the plutonium release from CWF is very low at less than 0.005 g/m{sup 2}.

Frank, S M [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Barber, T L [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Cummings, D G [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; DiSanto, T [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Esh, D W [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; Giglio, J J [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Goff, K M [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Johnson, S G [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Kennedy, J R [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Jue, J-F [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Noy, M [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; O'Holleran, T P [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Sinkler, W [UOP LLC, 25 E Algonquin Road, Des Plaines, IL 60017

2006-03-27T23:59:59.000Z

163

Leach tests of simulated low-level transuranic waste forms containing transuranic elements  

SciTech Connect (OSTI)

Simulations of waste forms that could be produced by slagging pyrolysis incineration of low-level transuranic (TRU) wastes stored at the Idaho National Engineering Laboratory (INEL) have been fabricated containing the transuranic isotopes /sup 237/Np, /sup 239/Pu, /sup 241/Am, and /sup 244/Cm at levels of approximately 1 ..mu..Ci/g of each. Leach tests were performed on frit; concrete monoliths made with frit and Portland cement; and vitrified monoliths of average INEL TRU waste, INEL soil, and simulated Rocky Flats plant sludge. Static leach tests were performed at 90, 70, 40, and 25/sup 0/C in deionized water for up to 364 days. Leachates were analyzed for the TRU elements by alpha spectrometry. From the leaching results the following generalizations can be made: (1) cemented frit and vitrified sludge waste forms produce leachates with the highest pHs (> 11) and have the lowest TRU leach rates, 10/sup -4/ g/m/sup 2/ d at 90/sup 0/C; (2) neptunium has a higher leach rate than the other three TRU elements by as much as two orders of magnitude for all waste forms tested except cemented frit; and (3) only the vitrified soil samples display a marked temperature dependence for leach rates of all four TRU elements.

Welch, J.M.; Sill, C.W.; Flinn, J.E.

1983-01-01T23:59:59.000Z

164

Leach tests of simulated low-level transuranic waste forms containing transuranic elements  

SciTech Connect (OSTI)

Simulations of waste forms that might be produced by slagging pyrolysis incineration of low-level transuranic (TRU) wastes stored at the Idaho National Engineering Laboratory (INEL) have been fabricated containing the transuranic isotopes /sup 237/Np, /sup 239/Pu, /sup 241/Am, /sup 244/Cm at levels of approximately 1 ..mu..Ci per gram of each. Leach tests were performed using frit and vitrified monolithic specimens of average INEL TRU waste, portland cement monoliths made with frit as aggregate, and vitrified monoliths of INEL soil and simulated Rocky Flats sludge. Static leach tests were performed at 90, 70, 40, and 25/sup 0/C in deionized water for up to 364 days. Leachates were analyzed for the TRU elements by alpha spectrometry. The following generalizations can be made: (1) Cemented frit and vitrified sludge waste forms produce leachates with the highest pHs (>11) and have the lowest TRU leach rates, 10/sup -4/ g/m/sup 2/.d at 90/sup 0/C. (2) Neptunium has a higher leach rate than the other three TRU elements by as much as two orders of magnitude for all waste forms tested except cemented frit. (3) Only the vitrified soil samples display a marked temperature dependence for leach rates of all four TRU elements.

Welch, J.M.; Sill, C.W.; Flinn, J.E.

1982-01-01T23:59:59.000Z

165

Final Environmental Impact Statement (Supplement to ERDA-1537, September 1977) Waste Management Operations Double-Shell Tanks for Defense High-Level Radioactive Waste Storage Savannah River Plant  

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

Do Do E/EIS-0062 FINAL ENVIRONMENTAL IMPACT mATEIUIENT (Supplement to ERDA-1537, September 1977) Waste ~ Management Operations Savannah River Plant ! Aiken, South Carolina Double-Shell Tanks for Defense High-Level Radioactive Waste Storage April 1980 U.S. DEPARTMENT OF ENERGY WASHINGTON. D.C.20545 1980 WL 94273 (F.R.) NOTICES DEPARTMENT OF ENERGY Office of Deputy Assistant Secretary for Nuclear Waste Management Double-Shell Tanks for Defense High-Level Radioactive Waste Storage, Savannah River Plant, Aiken, S.C. Wednesday, July 9, 1980 *46154 Record of Decision Decision. The decision has been made to complete the construction of the 14 double-shell tanks and use them to store defense high-level radioactive waste at the Savannah River Plant (SRP). Background. The SRP, located near Aiken, South Carolina, is a major installation of the

166

Idaho High-Level Waste & Facilities Disposition, Final Environmental Impact Statement  

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

Appendix A Appendix A Site Evaluation Process A-iii DOE/EIS-0287 Idaho HLW & FD EIS TABLE OF CONTENTS Section Page Appendix A Site Evaluation Process A-1 A.1 Introduction A-1 A.2 Methodology A-1 A.3 High-Level Waste Treatment and Interim Storage Site Selection A-3 A.3.1 Identification of "Must" Criteria A-3 A.3.2 Identification of "Want" Criteria A-3 A.3.3 Identification of Candidate Sites A-3 A.3.4 Evaluation Process A-4 A.3.5 Results of Evaluation Process A-6 A.4 Low-Activity Waste Disposal Site Selection A-6 A.4.1 Identification of "Must" Criteria A-7 A.4.2 Identification of "Want" Criteria A-8 A.4.3 Identification of Candidate Sites A-8 A.4.4 Evaluation Process A-8 A.4.5 Results of Evaluation Process A-9 A.4.6 Final Selection of a Low-Activity Waste Disposal Facility

167

Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment  

SciTech Connect (OSTI)

This data package documents the experimentally derived input data on the representative waste glasses; LAWA44, LAWB45, and LAWC22. This data will be used for Subsurface Transport Over Reactive Multi-phases (STORM) simulations of the Integrated Disposal Facility (IDF) for immobilized low-activity waste (ILAW). The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in July 2005. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali (Na+)-hydrogen (H+) ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow (PUF) and product consistency (PCT) tests where used for accelerated weathering or aging of the glasses in order to determine a chemical reaction network of secondary phases that form. The majority of the thermodynamic data used in this data package were extracted from the thermody-namic database package shipped with the geochemical code EQ3/6, version 8.0. Because of the expected importance of 129I release from secondary waste streams being sent to IDF from various thermal treatment processes, parameter estimates for diffusional release and solubility-controlled release from cementitious waste forms were estimated from the available literature.

Pierce, Eric M.; McGrail, B. Peter; Rodriguez, Elsa A.; Schaef, Herbert T.; Saripalli, Prasad; Serne, R. Jeffrey; Krupka, Kenneth M.; Martin, P. F.; Baum, Steven R.; Geiszler, Keith N.; Reed, Lunde R.; Shaw, Wendy J.

2004-09-01T23:59:59.000Z

168

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

SciTech Connect (OSTI)

Yucca Mountain, Nevada is currently being evaluated as a potential site for a geologic repository. As part of the repository assessment activities, candidate materials are being tested for possible use as construction materials for waste package containers. A large portion of this testing effort is focused on determining the long range corrosion properties, in a Yucca Mountain environment, for those materials being considered. Along similar lines, Argonne National Laboratory is testing a metallic alloy waste form that also is scheduled for disposal in a geologic repository, like Yucca Mountain. Due to the fact that Argonne`s waste form will require performance testing for an environment similar to what Yucca Mountain canister materials will require, this report was constructed to focus on the types of tests that have been conducted on candidate Yucca Mountain canister materials along with some of the results from these tests. Additionally, this report will discuss testing of Argonne`s metal waste form in light of the Yucca Mountain activities.

Keiser, D.D.

1996-11-01T23:59:59.000Z

169

Operational readiness review for the Waste Experimental Reduction Facility. Final report  

SciTech Connect (OSTI)

An Operational Readiness Review (ORR) at the Idaho National Engineering Laboratory`s (INEL`s) Waste Experimental Reduction Facility (WERF) was conducted by EG&G Idaho, Inc., to verify the readiness of WERF to resume operations following a shutdown and modification period of more than two years. It is the conclusion of the ORR Team that, pending satisfactory resolution of all pre-startup findings, WERF has achieved readiness to resume unrestricted operations within the approved safety basis. ORR appraisal forms are included in this report.

Not Available

1993-11-01T23:59:59.000Z

170

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

SciTech Connect (OSTI)

The U.S. Department of Energy Office of Environmental Management (EM) is engaging the national laboratories to provide the scientific and technological rigor to support EM program and project planning, technology development and deployment, project execution, and assessment of program outcomes. As an early demonstration of this new responsibility, Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) have been chartered to implement a science and technology program addressing low-temperature waste forms for immobilization of DOE aqueous waste streams, including technetium removal as an implementing technology. As a first step, the laboratories examined the technical risks and uncertainties associated with the Cast Stone waste immobilization projects at Hanford. Science and technology needs were identified for work associated with 1) conducting performance assessments and risk assessments of waste form and disposal system performance, and 2) technetium chemistry in tank wastes and separations of technetium from waste processing streams. Technical approaches to address the science and technology needs were identified and an initial sequencing priority was suggested. The following table summarizes the most significant science and technology needs and associated approaches to address those needs. These approaches and priorities will be further refined and developed as strong integrated teams of researchers from national laboratories, contractors, industry, and academia are brought together to provide the best science and technology solutions. Implementation of a science and technology program that addresses these needs by pursuing the identified approaches will have immediate benefits to DOE in reducing risks and uncertainties associated with near-term decisions regarding supplemental immobilization at Hanford. Longer term, the work has the potential for cost savings and for providing a strong technical foundation for future performance assessments at Hanford and across the DOE complex.

Fox, K.

2014-05-13T23:59:59.000Z

171

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

SciTech Connect (OSTI)

This Model Report describes the analysis and abstractions of the colloids process model for the waste form and engineered barrier system components of the total system performance assessment calculations to be performed with the Total System Performance Assessment-License Application model. Included in this report is a description of (1) the types and concentrations of colloids that could be generated in the waste package from degradation of waste forms and the corrosion of the waste package materials, (2) types and concentrations of colloids produced from the steel components of the repository and their potential role in radionuclide transport, and (3) types and concentrations of colloids present in natural waters in the vicinity of Yucca Mountain. Additionally, attachment/detachment characteristics and mechanisms of colloids anticipated in the repository are addressed and discussed. The abstraction of the process model is intended to capture the most important characteristics of radionuclide-colloid behavior for use in predicting the potential impact of colloid-facilitated radionuclide transport on repository performance.

R. Aguilar

2003-06-24T23:59:59.000Z

172

A summary of the sources of input parameter values for the Waste Isolation Pilot Plant final porosity surface calculations  

SciTech Connect (OSTI)

A summary of the input parameter values used in final predictions of closure and waste densification in the Waste Isolation Pilot Plant disposal room is presented, along with supporting references. These predictions are referred to as the final porosity surface data and will be used for WIPP performance calculations supporting the Compliance Certification Application to be submitted to the U.S. Environmental Protection Agency. The report includes tables and list all of the input parameter values, references citing their source, and in some cases references to more complete descriptions of considerations leading to the selection of values.

Butcher, B.M.

1997-08-01T23:59:59.000Z

173

Fabrication and Properties of Technetium-Bearing Pyrochlores and Perovskites as Potential Waste Forms - 13222  

SciTech Connect (OSTI)

Technetium-99 (t{sub 1/2}= 2.13x10{sup 5} years) is important from a nuclear waste perspective and is one of the most abundant, long-lived radioisotopes in used nuclear fuel (UNF). As such, it is targeted in UNF separation strategies such as UREX+, for isolation and encapsulation in solid waste forms for storage in a nuclear repository. We report here results regarding the incorporation of Tc-99 into ternary oxides of different structure types: pyrochlore (Nd{sub 2}Tc{sub 2}O{sub 7}), perovskite (SrTcO{sub 3}), and layered perovskite (Sr{sub 2}TcO{sub 4}). The goal was to determine synthesis conditions of these potential waste forms to immobilize Tc-99 as tetravalent technetium and to harvest crystallographic, thermophysical and hydrodynamic data. The objective of this research is to provide fundamental crystallographic and thermophysical data on advanced ceramic Tc-99 waste forms such as pyrochlore, perovskite, and layered perovskite in support of our current efforts on the corrosion of technetium-bearing waste forms. The ceramic Tc-99-bearing waste forms exhibit good crystallinity. The lattice parameters and crystal structures of the technetium host phases could be refined with high accuracies of ±3, ±4, and ±7 fm (10{sup -15} m), for Nd{sub 2}Tc{sub 2}O{sub 7}, SrTcO{sub 3}, and Sr{sub 2}TcO{sub 4}, respectively. The associated refinement residuals (R{sub Wp}) for the patterns are 4.1 %, 4.7 % and 6.7 %, and the refinement residuals for the individual phases (R{sub Bragg}) are 2.0 %, 2.4 % and 3.9 %, respectively. Thermophysical properties of the oxides SrTcO{sub 3}, Sr{sub 2}TcO{sub 4}, and Nd{sub 2}Tc{sub 2}O{sub 7} were analyzed using AC magnetic susceptibility measurements to further harvest information on the critical temperature (T{sub c}) for superconductivity. In our experiments the strontium technetates, SrTcO{sub 3} and Sr{sub 2}TcO{sub 4}, show superconductivity at rather high critical temperatures of T{sub c} = 7.8 K and 7 K, respectively. On the other hand Nd{sub 2}Tc{sub 2}O{sub 7} did not show any changes in magnetic properties above 3 K. (authors)

Hartmann, Thomas [University of Nevada - Las Vegas, Harry Reid Center, 4505 S. Maryland Parkway, Box 4009, Las Vegas, NV 89154-4009 (United States)] [University of Nevada - Las Vegas, Harry Reid Center, 4505 S. Maryland Parkway, Box 4009, Las Vegas, NV 89154-4009 (United States); Alaniz, Ariana J. [University of Nevada - Las Vegas, Howard R. Hughes College of Engineering, 4505 S. Maryland Parkway, Box 4009, Las Vegas, NV 89154-4009 (United States)] [University of Nevada - Las Vegas, Howard R. Hughes College of Engineering, 4505 S. Maryland Parkway, Box 4009, Las Vegas, NV 89154-4009 (United States); Antonio, Daniel J. [University of Nevada - Las Vegas, Department of Physics and Astronomy, 4505 S. Maryland Parkway, Box 4002, Las Vegas, NV 89154-4002 (United States)] [University of Nevada - Las Vegas, Department of Physics and Astronomy, 4505 S. Maryland Parkway, Box 4002, Las Vegas, NV 89154-4002 (United States)

2013-07-01T23:59:59.000Z

174

Final Hanford Site Solid (Radioactive and Hazardous) Waste Program Environmental Impact Statement Richland, Washington  

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

COVER SHEET COVER SHEET U.S. Department of Energy, Richland Operations Office TITLE: Final Hanford Site Solid (Radioactive and Hazardous) Waste Program Environmental Impact Statement, Richland, Benton County, Washington (DOE/EIS-0286F) CONTACT: For further information on this document, write or call: Mr. Michael S. Collins HSW EIS Document Manager Richland Operations Office U.S. Department of Energy, A6-38 P.O. Box 550 Richland, Washington 99352-0550 Telephone: (509) 376-6536 Fax: (509) 372-1926 Email: hsweis@rl.gov For further information on the Department's National Environmental Policy Act (NEPA) process, contact: Ms. Carol M. Borgstrom, Director Office of NEPA Policy and Compliance, EH-42 U.S. Department of Energy 1000 Independence Avenue, S.W.

175

Final Environmental Impact Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina  

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

I I I Y DoE/Els-o120 Final Environmental Impact Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina Volume 2 Q ~<$c'% ~ v ~ g ;:: # +4 -~ STATES O* December 1987 United States Department of Energy -- TABLE OF CONTENTS Appendix A GEOLOGY AND SUBSURFACE HYDROLOGY . . . . . . . . . . . . . . . A.1 Geology and Seismology . . . . . . . . . . . . . . . . . A.1.l Regional Geologic Setting . . . . . . . . . . . . A.1.1.1 Tectonic Provinces . . . . . . . . . . . A.I.1.2 Stratigraphy . . . . . . . . . . . . . . A.1.1.3 Geomorphology . . . . . . . . . . . . . . A.1.2 Seismology and Geologic Hazards . . . . . . . . . A.1.2.1 Geologic Structures and Seismicity . . . A.1.2.2 Seismic Events and Liquefaction Potentill . . . . . . . . . . . . . . . . A.2 Groundwater Resources . . . . . . . . . . . . . . . . . . A.2.1 Hydrostratigraphy . . . . . . . . . . . . . . . . A.2.2 Groundwater Hydrology . . . . . . . . . . . . . . A.2.2.1 Hydrologic Properties

176

Radiation damage of a glass-bonded zeolite waste form using ion irradiation.  

SciTech Connect (OSTI)

Glass-bonded zeolite is being considered as a candidate ceramic waste form for storing radioactive isotopes separated from spent nuclear fuel in the electrorefining process. To determine the stability of glass-bonded zeolite under irradiation, transmission electron microscope samples were irradiated using high energy helium, lead, and krypton. The major crystalline phase of the waste form, which retains alkaline and alkaline earth fission products, loses its long range order under both helium and krypton irradiation. The dose at which the long range crystalline structure is lost is about 0.4 dpa for helium and 0.1 dpa for krypton. Because the damage from lead is localized in such a small region of the sample, damage could not be recognized even at a peak damage of 50 dpa. Because the crystalline phase loses its long range structure due to irradiation, the effect on retention capacity needs to be further evaluated.

Allen, T. R.; Storey, B. G.

1997-12-05T23:59:59.000Z

177

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

SciTech Connect (OSTI)

Glass ceramics are being developed to immobilize fission products, separated from used nuclear fuel by aqueous reprocessing, into a stable waste form suitable for disposal in a geological repository. This work documents the glass ceramic formulation at bench scale and for a scaled melter test performed in a pilot-scale (approximately 1/4 scale) cold crucible induction melter (CCIM). Melt viscosity, electrical conductivity, and crystallization behavior upon cooling were measured on a small set of compositions to select a formulation for melter testing. Property measurements also identified a temperature range for melter operation and cooling profiles necessary to crystallize the targeted phases in the waste form. Bench scale and melter run results successfully demonstrate the processability of the glass ceramic using the CCIM melter technology.

Jarrod Crum [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Vince Maio [Idaho National Laboratory (INL), Idaho Falls, ID (United States); John McCloy [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Clark Scott [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Brian Riley [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Brad Benefiel [Idaho National Laboratory (INL), Idaho Falls, ID (United States); John Vienna [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Kip Archibald [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Carmen Rodriguez [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Veronica Rutledge [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zihua Zhu [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Joe Ryan [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Matthew Olszta [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

2014-01-01T23:59:59.000Z

178

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

DOE Patents [OSTI]

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

Colombo, P.; Kalb, P.D.

1984-06-05T23:59:59.000Z

179

A Science-Based Approach to Understanding Waste Form Durability in Open and Closed Nuclear Fuel Cycles  

SciTech Connect (OSTI)

There are two compelling reasons for understanding source term and near-field processes in a radioactive waste geologic repository. First, almost all of the radioactivity is initially in the waste form, mainly in the spent nuclear fuel (SNF) or nuclear waste glass. Second, over long periods, after the engineered barriers are degraded, the waste form is a primary control on the release of radioactivity. Thus, it is essential to know the physical and chemical state of the waste form after hundreds of thousands of years. The United States Department of Energy's Yucca Mountain Repository Program has initiated a long-term program to develop a basic understanding of the fundamental mechanisms of radionuclide release and a quantification of the release as repository conditions evolve over time. Specifically, the research program addresses four critical areas: (a) SNF dissolution mechanisms and rates; (b) formation and properties of U{sup 6+}-secondary phases; (c) waste form-waste package interactions in the near-field; and (d) integration of in-package chemical and physical processes. The ultimate goal is to integrate the scientific results into a larger scale model of source term and near-field processes. This integrated model will be used to provide a basis for understanding the behavior of the source term over long time periods (greater than 10{sup 5} years). Such a fundamental and integrated experimental and modeling approach to source term processes can also be readily applied to development of advanced waste forms as part of a closed nuclear fuel cycle. Specifically, a fundamental understanding of candidate waste form materials stability in high temperature/high radiation environments and near-field geochemical/hydrologic processes could enable development of advanced waste forms ''tailored'' to specific geologic settings.

M.T. Peters; R.C. Ewing

2006-06-22T23:59:59.000Z

180

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

SciTech Connect (OSTI)

The U.S. Department of Energy's (DOE) Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). In addition, the WTP LAW Vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as Cs-137, I-129, Tc-99, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750 C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product, which is one of the objectives of this current study, is being investigated to prevent dispersion during transport or burial/storage but is not necessary for performance. FBSR testing of a Hanford LAW simulant and a WTP-SW simulant at the pilot scale was performed by THOR Treatment Technologies, LLC at Hazen Research Inc. in April/May 2008. The Hanford LAW simulant was the Rassat 68 tank blend and the target concentrations for the LAW was increased by a factor of 10 for Sb, As, Ag, Cd, and Tl; 100 for Ba and Re (Tc surrogate); 1,000 for I; and 254,902 for Cs based on discussions with the DOE field office and the environmental regulators and an evaluation of the Hanford Tank Waste Envelopes A, B, and C. It was determined through the evaluation of the actual tank waste metals concentrations that some metal levels were not sufficient to achieve reliable detection in the off-gas sampling. Therefore, the identified metals concentrations were increased in the Rassat simulant processed by TTT at HRI to ensure detection and enable calculation of system removal efficiencies, product retention efficiencies, and mass balance closure without regard to potential results of those determinations or impacts on product durability response such as Toxicity Characteristic Leach Procedure (TCLP). A WTP-SW simulant based on melter off-gas analyses from Vitreous State Laboratory (VSL) was also tested at HRI in the 15-inch diameter Engineering Scale Test Demonstration (ESTD) dual reformer at HRI in 2008. The target concentrations for the Resource Conservation and Recovery Act (RCRA) metals were increased by 16X for Se, 29X for Tl, 42X for Ba, 48X for Sb, by 100X for Pb and Ni, 1000X for Ag, and 1297X for Cd to ensure detection by the an

Crawford, C.; Jantzen, C.

2012-02-02T23:59:59.000Z

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181

USING CENTER HOLE HEAT TRANSFER TO REDUCE FORMATION TIMES FOR CERAMIC WASTE FORMS FROM PYROPROCESSING  

SciTech Connect (OSTI)

The waste produced from processing spent fuel from the EBR II reactor must be processed into a waste form suitable for long term storage in Yucca Mountain. The method chosen produces zeolite granules mixed with glass frit, which must then be converted into a solid. This is accomplished by loading it into a can and heating to 900 C in a furnace regulated at 915 C. During heatup to 900 C, the zeolite and glass frit react and consolidate to produce a sodalite monolith. The resultant ceramic waste form (CWF) is then cooled. The waste is 52 cm in diameter and initially 300 cm long but consolidates to 150 cm long during the heating process. After cooling it is then inserted in a 5-DHLW/DOE SNF Long Canister. Without intervention, the waste takes 82 hours to heat up to 900 C in a furnace designed to geometrically fit the cylindrical waste form. This paper investigates the reduction in heating times possible with four different methods of additional heating through a center hole. The hole size is kept small to maximize the amount of CWF that is processed in a single run. A hole radius of 1.82 cm was selected which removes only 1% of the CWF. A reference computation was done with a specified inner hole surface temperature of 915 C to provide a benchmark for the amount of improvement which can be made. It showed that the heatup time can potentially be reduced to 43 hours with center hole heating. The first method, simply pouring high temperature liquid aluminum into the hole, did not produce any noticeable effect on reducing heat up times. The second method, flowing liquid aluminum through the hole, works well as long as the velocity is high enough (2.5 cm/sec) to prevent solidification of the aluminum during the initial front movement of the aluminum into the center hole. The velocity can be reduced to 1 cm/sec after the initial front has traversed the ceramic. This procedure reduces the formation time to near that of the reference case. The third method, flowing a gas through the center hole, also works well as long as the heat capacity times the velocity of the gas is equivalent to that of the flowing aluminum, and the velocity is high enough to produce an intermediate size heat transfer coefficient. The fourth method, using an electric heater, works well and heater sizes between 500 to 1000 Watts are adequate. These later three methods all can reduce the heatup time to 44 hours.

Kenneth J. Bateman; Charles W. Solbrig

2006-07-01T23:59:59.000Z

182

Management of solid wastes from the Limestone Injection Dry Scrubbing (LIDS) clean coal technology. Final report  

SciTech Connect (OSTI)

The objectives of this project were to characterize by-products from a pilot Limestone Injection Dry Scrubbing (LIDS) process and to develop processes directed toward the safe and economic use or disposal of these wastes. Because LIDS is a developing Clean Coal technology, a database of chemical and physical characteristics of the by-product was first developed. During the course of this project, it was found that the waste alone did not form high-strength products sufficient for use in construction and engineering applications. Therefore, the project was redirected to evaluate the by-product as a soil-cement and Portland cement raw material, agricultural liming agent, backfill/landfill material component, and mine reclamation/neutralizing agent. Based on these evaluations, the most viable uses for the LIDS byproduct include use in mine reclamation or as a neutralization agent. If soluble sulfites can be minimized by avoiding a dolomitic LIDS reagent, use as an agricultural liming agent has promise. Interest from an Ohio utility in the LIDS process suggests possible application of results at the demonstration or commercial stages.

Musiol, W.F. Jr.; Czuczwa, J.M.

1993-03-01T23:59:59.000Z

183

Town of Hague landfill reclamation study: Research ways to increase waste heating value and reduce waste volume. Final report  

SciTech Connect (OSTI)

Monitored composing was studied as a method for reducing the quantity of waste requiring disposed from a landfill reclamation project. After each of two re-screening steps, composted {open_quotes}soil{close_quotes} from a single long windrow of varying depths and moisture content was subjected to analytical testing to determine its suitability to remain as backfill in a reclaimed landfill site. The remaining uncomposted waste was combusted at a waste-to-energy facility to determine if Btu values were improved. Results indicate that a full-scale composting operation could result in a net decrease of approximately 11 percent in disposal costs. The Btu value of the reclaimed waste was calculated to be 4,500 to 5,000 Btu/lb. The feasibility of composting reclaimed waste at other landfill reclamation projects will depend upon site-specific technical and economic factors, including size and nature of the organic fraction of the waste mass, local processing costs, and the cost of waste disposal alternatives.

Salerni, E. [SSB Environmental Inc., Albany, NY (United States)

1997-01-01T23:59:59.000Z

184

Underground waste barrier structure  

DOE Patents [OSTI]

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

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

1988-01-01T23:59:59.000Z

185

In situ vitrification application to buried waste: Final report of intermediate field tests at Idaho National Engineering Laboratory  

SciTech Connect (OSTI)

This report describes two in situ vitrification field tests conducted on simulated buried waste pits during June and July 1990 at the Idaho National Engineering Laboratory. In situ vitrification, an emerging technology for in place conversion of contaminated soils into a durable glass and crystalline waste form, is being investigated as a potential remediation technology for buried waste. The overall objective of the two tests was to access the general suitability of the process to remediate waste structures representative of buried waste found at Idaho National Engineering Laboratory. In particular, these tests, as part of a treatability study, were designed to provide essential information on the field performance of the process under conditions of significant combustible and metal wastes and to test a newly developed electrode feed technology. The tests were successfully completed, and the electrode feed technology successfully processed the high metal content waste. Test results indicate the process is a feasible technology for application to buried waste. 33 refs., 109 figs., 39 tabs.

Callow, R.A.; Weidner, J.R.; Loehr, C.A.; Bates, S.O. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Thompson, L.E.; McGrail, B.P. (Pacific Northwest Lab., Richland, WA (United States))

1991-08-01T23:59:59.000Z

186

Compliance with Waste Acceptance Criteria of WIPP and NTS for Vitrified Low-Level and TRU Waste Forms  

SciTech Connect (OSTI)

A joint project between the Oak Ridge National Laboratory (ORNL) and the Savannah River Technology Center (SRTC) has been established to evaluate vitrification as an option for the immobilization of waste within ORNL tank farms. This paper presents details of calculations based on current best available analyses of the Oak Ridge Tanks on the limits for waste loadings imposed by the waste acceptance criteria.

Harbour, J.R. [Westinghouse Savannah River Company, AIKEN, SC (United States); Andrews, M.K.

1998-07-01T23:59:59.000Z

187

Microsoft Word - Tran Waste final report 2-8-05.doc  

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

Transuranic Waste Management at Transuranic Waste Management at Los Alamos National Laboratory DOE/IG-0673 February 2005 REPORT ON TRANSURANIC WASTE MANAGEMENT AT LOS ALAMOS NATIONAL LABORATORY TABLE OF CONTENTS Legacy Transuranic Waste Disposal Details of Finding 1 Recommendations and Comments 4 Appendices 1. Objective, Scope, and Methodology 6 2. Transuranic Waste Storage 8 3. Prior Audit Reports 10 4. Management Comments 11 Legacy Transuranic Waste Disposal Page 1 Details of Finding Background Los Alamos National Laboratory (Los Alamos) has been involved in the development, production, and maintenance of the Nation's nuclear weapons stockpile for over six decades. Research, development, and fabrication of weapons components produced a

188

The application of a chemical equilibrium model, SOLTEQ, to predict the chemical speciations in stabilized/solidified waste forms  

E-Print Network [OSTI]

THE APPLICATION OI' A CHEMICAL EQUILIBRIUM MODEL, SOLTEQ, TO PREDICT THK CHEMICAL SPKCIATIONS IN STABILIZED/SOLIDIFIED WASTE FORMS A Thesis by JOO-YANG PARK Submitted to the Office of Graduate Studies of Texas A&M University in partial... fulfillment of the requirements for the degree of MASTER OF SCIENCE December 1994 Major Subject: Civil Engineering THE APPLICATION OF A CHEMICAL EQUILIBRIUM MODEL, SOLTEQ, TO PREDICT THE CHEMICAL SPECIATIONS IN STABILIZED/SOLIDIFIED WASTE FORMS A Thesis...

Park, Joo-Yang

1994-01-01T23:59:59.000Z

189

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

SciTech Connect (OSTI)

This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system, and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity. (authors)

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

2013-07-01T23:59:59.000Z

190

Fluidized-bed waste-heat recovery system development: Final report  

SciTech Connect (OSTI)

A major energy loss in industry is the heat content of the flue gases from industrial process heaters. One effective way to utilize the energy, which is applicable to all processes, is to preheat the combustion air for the process heater. Although recuperators are available to preheat this air when the flue gases are clean, recuperators to recover the heat from dirty and corrosive flue gases do not exist. The Fluidized-Bed Waste-Heat Recovery (FBWHR) system is designed to preheat this combustion air using the heat available in dirty flue gas streams. In this system, recirculating alumina particles are heated by the flue gas in a raining bed. The hot particles are then removed from the bed and placed in a fluidized bed where they are fluidized by the combustion air. Through this process, the combustion air is preheated. The cooled particles are then returned to the raining bed. Initial development of this concept is for the aluminum smelting industry. In this final report, the design, development, fabrication, and installation of a full-scale FBWHR system is detailed.

Patch, K.D.; Cole, W.E.

1988-06-01T23:59:59.000Z

191

Method and apparatus for using hazardous waste form non-hazardous aggregate  

SciTech Connect (OSTI)

This patent describes an apparatus for converting hazardous waste into non-hazardous, non-leaching aggregate, the apparatus. It comprises: a source of particulate solid materials, volatile gases and gaseous combustion by-products; oxidizing means comprising at least one refractory-lined, water-cooled, metal-walled vessel; means for introducing the particulate solid material, volatile gases and gaseous combustion by-products to the oxidizing means; means for inducing combustion in the oxidizing means, the heat of combustion forming molten slag and noncombustible fines from noncombustible material; means for accumulating the slag; means for introducing the noncombustible fines to the molten slag; means for removing the mixture from the apparatus; and means for cooling the mixture to form the non-hazardous, non-leaching aggregates.

Kent, J.M.; Robards, H.L. Jr.

1992-07-28T23:59:59.000Z

192

Final Environmental Assessment for Waste Disposition Activities at the Paducah Site Paducah, Kentucky  

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

0-347(doc)/093002 0-347(doc)/093002 1 FINDING OF NO SIGNIFICANT IMPACT WASTE DISPOSITION ACTIVITIES AT THE PADUCAH SITE PADUCAH, KENTUCKY AGENCY: U.S. DEPARTMENT OF ENERGY ACTION: FINDING OF NO SIGNIFICANT IMPACT SUMMARY: The U.S. Department of Energy (DOE) has completed an environmental assessment (DOE/EA-1339), which is incorporated herein by reference, for proposed disposition of polychlorinated biphenyl (PCB) wastes, low-level radioactive waste (LLW), mixed low- level radioactive waste (MLLW), and transuranic (TRU) waste from the Paducah Gaseous Diffusion Plant Site (Paducah Site) in Paducah, Kentucky. All of the wastes would be transported for disposal at various locations in the United States. Based on the results of the impact analysis reported in the EA, DOE has determined that the proposed action is

193

Waste Isolation Pilot Plant Title I operator dose calculations. Final report, LATA report No. 90  

SciTech Connect (OSTI)

The radiation exposure dose was estimated for the Waste Isolation Pilot Plant (WIPP) operating personnel who do the unloading and transporting of the transuranic contact-handled waste. Estimates of the radiation source terms for typical TRU contact-handled waste were based on known composition and properties of the waste. The operations sequence for waste movement and storage in the repository was based upon the WIPP Title I data package. Previous calculations had been based on Conceptual Design Report data. A time and motion sequence was developed for personnel performing the waste handling operations both above and below ground. Radiation exposure calculations were then performed in several fixed geometries and folded with the time and motion studies for individual workers in order to determine worker exposure on an annual basis.

Hughes, P.S.; Rigdon, L.D.

1980-02-01T23:59:59.000Z

194

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

SciTech Connect (OSTI)

One of the immobilization technologies under consideration as a Supplemental Treatment for Hanford’s Low Activity Waste (LAW) is Fluidized Bed Steam Reforming (FBSR). The FBSR technology forms a mineral waste form at moderate processing temperatures thus retaining and atomically bonding the halides, sulfates, and technetium in the mineral phases (nepheline, sodalite, nosean, carnegieite). Additions of kaolin clay are used instead of glass formers and the minerals formed by the FBSR technology offers (1) atomic bonding of the radionuclides and constituents of concern (COC) comparable to glass, (2) short and long term durability comparable to glass, (3) disposal volumes comparable to glass, and (4) higher Na2O and SO{sub 4} waste loadings than glass. The higher FBSR Na{sub 2}O and SO{sub 4} waste loadings contribute to the low disposal volumes but also provide for more rapid processing of the LAW. Recent FBSR processing and testing of Hanford radioactive LAW (Tank SX-105 and AN-103) waste is reported and compared to previous radioactive and non-radioactive LAW processing and testing.

Jantzen, Carol; Herman, Connie; Crawford, Charles; Bannochie, Christopher; Burket, Paul; Daniel, Gene; Cozzi, Alex; Nash, Charles; Miller, Donald; Missimer, David

2014-01-10T23:59:59.000Z

195

Program Final Report - Develop Thermoelectric Technology for Automotive Waste Heat Recovery  

SciTech Connect (OSTI)

We conducted a vehicle analysis to assess the feasibility of thermoelectric technology for waste heat recovery and conversion to useful electrical power and found that eliminating the 500 W of electrical power generated by the alternator corresponded to about a 7% increase in fuel economy (FE) for a small car and about 6% for a full size truck. Electric power targets of 300 W were established for city and highway driving cycles for this project. We obtained critical vehicle level information for these driving cycles that enabled a high-level design and performance analysis of radiator and exhaust gas thermoelectric subsystems for several potential vehicle platforms, and we identified the location and geometric envelopes of the radiator and exhaust gas thermoelectric subsystems. Based on this analysis, we selected the Chevrolet Suburban as the most suitable demonstration vehicle for this project. Our modeling and thermal analysis assessment of a radiator-based thermoelectric generator (TEG), however, revealed severe practical limitations. Specifically the small temperature difference of 100°C or less between the engine coolant and ambient air results in a low Carnot conversion efficiency, and thermal resistance associated with air convection would reduce this conversion efficiency even further. We therefore decided not to pursue a radiator-based waste heat recovery system and focused only on the exhaust gas. Our overall approach was to combine science and engineering: (1) existing and newly developed TE materials were carefully selected and characterized by the material researcher members of our team, and most of the material property results were validated by our research partners, and (2) system engineers worked closely with vehicle engineers to ensure that accurate vehicle-level information was used for developing subsystem models and designs, and the subsystem output was analyzed for potential fuel economy gains. We incorporated material, module, subsystem, and integration costs into the material selection criteria in order to balance various materials, module and subsystem design, and vehicle integration options. Our work on advanced TE materials development and on TEG system design, assembly, vehicle integration, and testing proceeded in parallel efforts. Results from our two preliminary prototype TEGs using only Bi-Te TE modules allowed us to solve various mechanical challenges and to finalize and fine tune aspects of the design and implementation. Our materials research effort led us to quickly abandon work on PbTe and focus on the skutterudite materials due to their superior mechanical performance and suitability at automotive exhaust gas operating temperatures. We synthesized a sufficiently large quantity of skutterudite material for module fabrication for our third and final prototype. Our TEG#3 is the first of its kind to contain state-of-the-art skutterudite-based TE modules to be installed and tested on a production vehicle. The design, which consisted of 24 skutterudite modules and 18 Bi-Te modules, attempted to optimize electrical power generation by using these two kinds of TE modules that have their peak performance temperatures matched to the actual temperature profile of the TEG during operation. The performance of TEG#3 was limited by the maximum temperature allowable for the Bi-Te TE modules located in the colder end of the TEG, resulting in the operating temperature for the skutterudite modules to be considerably below optimum. We measured the power output for (1) the complete TEG (25 Watts) and (2) an individual TE module series string (1/3 of the TEG) operated at a 60°C higher temperature (19 Watts). We estimate that under optimum operating temperature conditions, TEG#3 will generate about 235 Watts. With additional improvements in thermal and electrical interfaces, temperature homogeneity, and power conditioning, we estimate TEG#3 could deliver a power output of about 425 Watts.

Gregory Meisner

2011-08-31T23:59:59.000Z

196

Final environmental impact statement. Management of commercially generated radioactive waste. Volume 2. Appendices  

SciTech Connect (OSTI)

This EIS analyzes the significant environmental impacts that could occur if various technologies for management and disposal of high-level and transuranic wastes from commercial nuclear power reactors were to be developed and implemented. This EIS will serve as the environmental input for the decision on which technology, or technologies, will be emphasized in further research and development activities in the commercial waste management program. The action proposed in this EIS is to (1) adopt a national strategy to develop mined geologic repositories for disposal of commercially generated high-level and transuranic radioactive waste (while continuing to examine subseabed and very deep hole disposal as potential backup technologies) and (2) conduct a R and D program to develop such facilities and the necessary technology to ensure the safe long-term containment and isolation of these wastes. The Department has considered in this statement: development of conventionally mined deep geologic repositories for disposal of spent fuel from nuclear power reactors and/or radioactive fuel reprocessing wastes; balanced development of several alternative disposal methods; and no waste disposal action. This volume contains appendices of supplementary data on waste management systems, geologic disposal, radiological standards, radiation dose calculation models, related health effects, baseline ecology, socio-economic conditions, hazard indices, comparison of defense and commercial wastes, design considerations, and wastes from thorium-based fuel cycle alternatives. (DMC)

Not Available

1980-10-01T23:59:59.000Z

197

Round-robin testing of a reference glass for low-activity waste forms  

SciTech Connect (OSTI)

A round robin test program was conducted with a glass that was developed for use as a standard test material for acceptance testing of low-activity waste glasses made with Hanford tank wastes. The glass is referred to as the low-activity test reference material (LRM). The program was conducted to measure the interlaboratory reproducibility of composition analysis and durability test results. Participants were allowed to select the methods used to analyze the glass composition. The durability tests closely followed the Product Consistency Test (PCT) Method A, except that tests were conducted at both 40 and 90 C and that parallel tests with a reference glass were not required. Samples of LRM glass that had been crushed, sieved, and washed to remove fines were provided to participants for tests and analyses. The reproducibility of both the composition and PCT results compare favorably with the results of interlaboratory studies conducted with other glasses. From the perspective of reproducibility of analysis results, this glass is acceptable for use as a composition standard for nonradioactive components of low-activity waste forms present at >0.1 elemental mass % and as a test standard for PCTS at 40 and 90 C. For PCT with LRM glass, the expected test results at the 95% confidence level are as follows: (1) at 40 C: pH = 9.86 {+-} 0.96; [B] = 2.30 {+-} 1.25 mg/L; [Na] = 19.7 {+-} 7.3 mg/L; [Si] = 13.7 {+-} 4.2 mg/L; and (2) at 90 C: pH = 10.92 {+-} 0.43; [B] = 26.7 {+-} 7.2 mg/L; [Na] = 160 {+-} 13 mg/L; [Si] = 82.0 {+-} 12.7 mg/L. These ranges can be used to evaluate the accuracy of PCTS conducted at other laboratories.

Ebert, W. L.; Wolf, S. F.

1999-12-06T23:59:59.000Z

198

Final Environmental Impact Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina  

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

PURPOSE PURPOSE The U.S. Department of Energv SUMRY (DOE) has Dreuared this environmental impact -. . . statement (EIS) to assess the environmental consequences of the implementation of modified waste management activities for hazardous, low-level radioactive, and mixed wastes for the protection of groundwater, human health, and the environment at its Savannah River Plant (SRP) in Aiken, South Carolina. This EIS, which is both programmatic and project-specific, has been prepared in accordance with Section 102(2)(C) of the National Environmental Policy Act (NEPA) of 1969, as amended. It is intended to support broad decisions on future actions on SRP waste management activities and to provide project- related environmental input and support for project-specific decisions on pro- ceeding with cleanup activities at existing waste sites in the R- and F-Areas, establishing new waste

199

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

SciTech Connect (OSTI)

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

Not Available

1980-08-01T23:59:59.000Z

200

Initial Evaluation of Processing Methods for an Epsilon Metal Waste Form  

SciTech Connect (OSTI)

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

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

2012-06-11T23:59:59.000Z

Note: This page contains sample records for the topic "final waste form" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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201

Final report of the Department of Energy pilot internship program on radioactive waste at Vanderbilt University (September 1, 1993-08/31, 1994)  

SciTech Connect (OSTI)

This final report summarizes Vanderbilt's ten year program in radioactive waste management. The report describes the interns selected for the program, the interns' course of study, and their assignments.

Frank Parker

1999-08-31T23:59:59.000Z

202

Impact of DOE Orders on Waste Management Operation. Volume 1: [Final report  

SciTech Connect (OSTI)

Department of Energy Orders are the internal documents which govern the management of all Department of Energy facilities and operations. DOE Orders are the vehicles by which Federal and state laws and regulations are applied to Department of Energy activities. A selected set of 22 Department of Energy Orders was reviewed to identify the applicability and impact of each Order on waste management operations at Los Alamos National Laboratory. Of the 22 Orders reviewed, five set forth requirements which have a high degree of impact on waste management activities. Eight Orders have a moderate degree of impact on waste management activities, and the remaining nine Orders have a low degree of impact.

Klein, R.B.; Jennrich, E.A.; Shuman, R.; Sandquist, G.M. [Rogers and Associates Engineering Corp., Salt Lake City, UT (United States); Rutz, A.C.; Littleton, M.W.; McCauce, C.H. [Wastren, Inc., Idaho Falls, ID (United States)

1989-09-01T23:59:59.000Z

203

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

SciTech Connect (OSTI)

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

Ehst, D.; Nuclear Engineering Division

2010-08-04T23:59:59.000Z

204

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

SciTech Connect (OSTI)

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

TOMASZEWSKI, T.A.

2000-04-25T23:59:59.000Z

205

Infrastructure support for a waste management institute. Final project report, September 12, 1994--September 11, 1997  

SciTech Connect (OSTI)

North Carolina A and T State University has completed the development of an infrastructure for the interdisciplinary Waste Management Institute (WMI). The Interdisciplinary Waste Management Institute (WMI) was approved in June, 1994 by the General Administration of the University of North Carolina as an academic support unit with research and public service functions. The mission of the WMI is to enhance awareness and understanding of waste management issues and to provide instructional support including research and outreach. The goals of WMI are as follows: increase the number of minority professionals who will work in waste management fields; develop cooperative and exchange programs involving faculty, students, government, and industry; serve as institutional sponsor of public awareness workshops and lecture series; and support interdisciplinary research programs. The vision of the WMI is to provide continued state-of-the art environmental educational programs, research, and outreach.

NONE

1997-11-01T23:59:59.000Z

206

FINAL  

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

2 2 FINAL ENVIRONMENTAL ASSESSMENT FOR EXIDE TECHNOLOGIES ELECTRIC DRIVE VEHICLE BATTERY AND COMPONENT MANUFACTURING INITIATIVE APPLICATION, BRISTOL, TN, AND COLUMBUS, GA U.S. Department of Energy National Energy Technology Laboratory March 2010 DOE/EA-1712 FINAL ENVIRONMENTAL ASSESSMENT FOR EXIDE TECHNOLOGIES ELECTRIC DRIVE VEHICLE BATTERY AND COMPONENT MANUFACTURING INITIATIVE APPLICATION, BRISTOL, TN, AND COLUMBUS, GA U.S. Department of Energy National Energy Technology Laboratory March 2010 DOE/EA-1712 iii COVER SHEET Responsible Agency: U.S. Department of Energy (DOE) Title: Environmental Assessment for Exide Technologies Electric Drive Vehicle Battery and Component Manufacturing Initiative Application, Bristol, TN, and Columbus, GA

207

Effect of glass composition on waste form durability: A critical review  

SciTech Connect (OSTI)

This report reviews literature concerning the relationship between the composition and durability of silicate glasses, particularly glasses proposed for immobilization of radioactive waste. Standard procedures used to perform durability tests are reviewed. It is shown that tests in which a low-surface area sample is brought into contact with a very large volume of solution provide the most accurate measure of the intrinsic durability of a glass composition, whereas high-surface area/low-solution volume tests are a better measure of the response of a glass to changes in solution chemistry induced by a buildup of glass corrosion products. The structural chemistry of silicate and borosilicate glasses is reviewed to identify those components with the strongest cation-anion bonds. A number of examples are discussed in which two or more cations engage in mutual bonding interactions that result in minima or maxima in the rheologic and thermodynamic properties of the glasses at or near particular optimal compositions. It is shown that in simple glass-forming systems such interactions generally enhance the durability of glasses. Moreover, it is shown that experimental results obtained for simple systems can be used to account for durability rankings of much more complex waste glass compositions. Models that purport to predict the rate of corrosion of glasses in short-term durability tests are evaluated using a database of short-term durability test results for a large set of glass compositions. The predictions of these models correlate with the measured durabilities of the glasses when considered in large groupings, but no model evaluated in this review provides accurate estimates of durability for individual glass compositions. Use of these models in long-term durability models is discussed. 230 refs.

Ellison, A.J.G.; Mazer, J.J.; Ebert, W.L. [Argonne National Lab., IL (United States). Chemical Technology Div.

1994-11-01T23:59:59.000Z

208

Combustion of municipal solid wastes with oil shale in a circulating fluidized bed. Final report  

SciTech Connect (OSTI)

The problem addressed by our invention is that of municipal solid waste utilization. The dimensions of the problem can be visualized by the common comparison that the average individual in America creates in five years time an amount of solid waste equivalent in weight to the Statue of Liberty. The combustible portion of the more than 11 billion tons of solid waste (including municipal solid waste) produced in the United States each year, if converted into useful energy, could provide 32 quads per year of badly needed domestic energy, or more than one-third of our annual energy consumption. Conversion efficiency and many other factors make such a production level unrealistic, but it is clear that we are dealing with a very significant potential resource. This report describes research pertaining to the co-combustion of oil shale with solid municipal wastes in a circulating fluidized bed. The oil shale adds significant fuel content and also constituents that can possible produce a useful cementitious ash.

NONE

1996-06-30T23:59:59.000Z

209

Waste Isolation Pilot Plant Materials Interface Interactions Test: Papers presented at the Commission of European Communities workshop on in situ testing of radioactive waste forms and engineered barriers  

SciTech Connect (OSTI)

The three papers in this report were presented at the second international workshop to feature the Waste Isolation Pilot Plant (WIPP) Materials Interface Interactions Test (MIIT). This Workshop on In Situ Tests on Radioactive Waste Forms and Engineered Barriers was held in Corsendonk, Belgium, on October 13--16, 1992, and was sponsored by the Commission of the European Communities (CEC). The Studiecentrum voor Kernenergie/Centre D`Energie Nucleaire (SCK/CEN, Belgium), and the US Department of Energy (via Savannah River) also cosponsored this workshop. Workshop participants from Belgium, France, Germany, Sweden, and the United States gathered to discuss the status, results and overviews of the MIIT program. Nine of the twenty-five total workshop papers were presented on the status and results from the WIPP MIIT program after the five-year in situ conclusion of the program. The total number of published MIIT papers is now up to almost forty. Posttest laboratory analyses are still in progress at multiple participating laboratories. The first MIIT paper in this document, by Wicks and Molecke, provides an overview of the entire test program and focuses on the waste form samples. The second paper, by Molecke and Wicks, concentrates on technical details and repository relevant observations on the in situ conduct, sampling, and termination operations of the MIIT. The third paper, by Sorensen and Molecke, presents and summarizes the available laboratory, posttest corrosion data and results for all of the candidate waste container or overpack metal specimens included in the MIIT program.

Molecke, M.A.; Sorensen, N.R. [eds.] [Sandia National Labs., Albuquerque, NM (US); Wicks, G.G. [ed.] [Westinghouse Savannah River Technology Center, Aiken, SC (US)

1993-08-01T23:59:59.000Z

210

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

SciTech Connect (OSTI)

The United States is now re-assessing its nuclear waste disposal policy and re-evaluating the option of moving away from the current once-through open fuel cycle to a closed fuel cycle. In a closed fuel cycle, used fuels will be reprocessed and useful components such as uranium or transuranics will be recovered for reuse. During this process, a variety of waste streams will be generated. Immobilizing these waste streams into appropriate waste forms for either interim storage or long-term disposal is technically challenging. Highly volatile or soluble radionuclides such as iodine ({sup 129}I) and technetium ({sup 99}Tc) are particularly problematic, because both have long half-lives and can exist as gaseous or anionic species that are highly soluble and poorly sorbed by natural materials. Under the support of Sandia National Laboratories (SNL) Laboratory-Directed Research & Development (LDRD), we have developed a suite of inorganic nanocomposite materials (SNL-NCP) that can effectively entrap various radionuclides, especially for {sup 129}I and {sup 99}Tc. In particular, these materials have high sorption capabilities for iodine gas. After the sorption of radionuclides, these materials can be directly converted into nanostructured waste forms. This new generation of waste forms incorporates radionuclides as nano-scale inclusions in a host matrix and thus effectively relaxes the constraint of crystal structure on waste loadings. Therefore, the new waste forms have an unprecedented flexibility to accommodate a wide range of radionuclides with high waste loadings and low leaching rates. Specifically, we have developed a general route for synthesizing nanoporous metal oxides from inexpensive inorganic precursors. More than 300 materials have been synthesized and characterized with x-ray diffraction (XRD), BET surface area measurements, and transmission electron microscope (TEM). The sorption capabilities of the synthesized materials have been quantified by using stable isotopes I and Re as analogs to {sup 129}I and {sup 99}Tc. The results have confirmed our original finding that nanoporous Al oxide and its derivatives have high I sorption capabilities due to the combined effects of surface chemistry and nanopore confinement. We have developed a suite of techniques for the fixation of radionuclides in metal oxide nanopores. The key to this fixation is to chemically convert a target radionuclide into a less volatile or soluble form. We have developed a technique to convert a radionuclide-loaded nanoporous material into a durable glass-ceramic waste form through calcination. We have shown that mixing a radionuclide-loaded getter material with a Na-silicate solution can effectively seal the nanopores in the material, thus enhancing radionuclide retention during waste form formation. Our leaching tests have demonstrated the existence of an optimal vitrification temperature for the enhancement of waste form durability. Our work also indicates that silver may not be needed for I immobilization and encapsulation.

Rodriguez, Mark Andrew; Bencoe, Denise Nora; Brinker, C. Jeffrey; Murphy, Andrew Wilson; Holt, Kathleen Caroline; Turnham, Rigney; Kruichak, Jessica Nicole; Tellez, Hernesto; Miller, Andy; Xiong, Yongliang; Pohl, Phillip Isabio; Ockwig, Nathan W.; Wang, Yifeng; Gao, Huizhen

2010-09-01T23:59:59.000Z

211

High-level waste canister storage final design, installation, and testing. Topical report  

SciTech Connect (OSTI)

This report is a description of the West Valley Demonstration Project`s radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project`s vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access.

Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M.; Vlad, P.M.

1998-04-01T23:59:59.000Z

212

A Research Needs Assessment for waste plastics recycling: Volume 1, Executive summary. Final report  

SciTech Connect (OSTI)

This first volume provides a summary of the entire project. The study utilized the talents of a large number of participants, including a significant number of peer reviewers from industrial companies, government agencies, and research institutes. in addition, an extensive analysis of relevant literature was carried out. In considering the attractiveness of recycling technologies that are alternatives to waste-to-energy combustion units, a systems approach was utilized. Collection of waste streams containing plastics, sortation, and reclamation of plastics and plastic mixtures, reprocessing or chemical conversion of the reclaimed polymers, and the applicability of the products to specific market segments have been analyzed in the study.

NONE

1994-12-01T23:59:59.000Z

213

The largest radioactive waste glassification  

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

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

214

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

SciTech Connect (OSTI)

The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to summarize the data collection in support of future melter demonstration testing for crystalline ceramic waste forms. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. The principal difficulties encountered during processing of the ?reference ceramic? waste form by a melt and crystallization process were the incomplete incorporation of Cs into the hollandite phase and the presence of secondary Cs-Mo non-durable phases. In the single phase hollandite system, these issues were addressed in this study by refining the compositions to include Cr as a transition metal element and the use of Ti/TiO{sub 2} buffer to maintain reducing conditions. Initial viscosity studies of ceramic waste forms indicated that the pour spout must be maintained above 1400{deg}C to avoid flow blockages due to crystallization. In-situ electron irradiations simulate radiolysis effects indicated hollandite undergoes a crystalline to amorphous transition after a radiation dose of 10{sup 13} Gy which corresponds to approximately 1000 years at anticipated doses (2?10{sup 10}-2?10{sup 11} Gy). Dual-beam ion irradiations employing light ion beam (such as 5 MeV alpha) and heavy ion beam (such as 100 keV Kr) studies indicate that reference ceramic waste forms are radiation tolerant to the ??particles and ?-particles, but are susceptible to a crystalline to amorphous transition under recoil nuclei effects. A path forward for refining the processing steps needed to form the targeted phase assemblages is outlined in this report. Processing modifications including melting in a reducing atmosphere with the use of Ti/TiO2 buffers, and the addition of Cr to the transition metal additives to facilitate Cs-incorporation in the hollandite phase. In addition to melt processing, alternative fabrication routes are being considered including Spark Plasma Sintering (SPS) and Hot Isostatic Pressing (HIP).

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

2012-09-21T23:59:59.000Z

215

Final Report. Conversion of Low Temperature Waste Heat Utilizing Hermetic Organic Rankine Cycle  

SciTech Connect (OSTI)

The design of waste heat recovery using the organic Rankine cycle (ORC) engine is updated. Advances in power electronics with lower cost enable the use of a single shaft, high-speed generator eliminating wear items and allowing hermetic sealing of the working fluid. This allows maintenance free operation and a compact configuration that lowers cost, enabling new market opportunities.

Fuller, Robert L.

2005-04-20T23:59:59.000Z

216

HANFORD MEDIUM-LOW CURIE WASTE PRETREATMENT ALTERNATIVES PROJECT FRACTIONAL CRYSTALLIZATION PILOT SCALE TESTING FINAL REPORT  

SciTech Connect (OSTI)

The Fractional Crystallization Pilot Plant was designed and constructed to demonstrate that fractional crystallization is a viable way to separate the high-level and low-activity radioactive waste streams from retrieved Hanford single-shell tank saltcake. The focus of this report is to review the design, construction, and testing details of the fractional crystallization pilot plant not previously disseminated.

HERTING DL

2008-09-16T23:59:59.000Z

217

Integrity assessment plan for PNL 300 area radioactive hazardous waste tank system. Final report  

SciTech Connect (OSTI)

The Pacific Northwest Laboratory (PNL), operated by Battelle Memorial Institute under contract to the U.S. Department of Energy, operates tank systems for the U.S. Department of Energy, Richland Operations Office (DOE-RL), that contain dangerous waste constituents as defined by Washington State Department of Ecology (WDOE) Dangerous Waste Regulations, Washington Administrative Code (WAC) 173-303-040(18). Chapter 173-303-640(2) of the WAC requires the performance of integrity assessments for each existing tank system that treats or stores dangerous waste, except those operating under interim status with compliant secondary containment. This Integrity Assessment Plan (IAP) identifies all tasks that will be performed during the integrity assessment of the PNL-operated Radioactive Liquid Waste Systems (RLWS) associated with the 324 and 325 Buildings located in the 300 Area of the Hanford Site. It describes the inspections, tests, and analyses required to assess the integrity of the PNL RLWS (tanks, ancillary equipment, and secondary containment) and provides sufficient information for adequate budgeting and control of the assessment program. It also provides necessary information to permit the Independent, Qualified, Registered Professional Engineer (IQRPE) to approve the integrity assessment program.

NONE

1996-03-01T23:59:59.000Z

218

Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group.  

SciTech Connect (OSTI)

In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD&D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years.

Collins, E.; Duguid, J.; Henry, R.; Karell, E.; Laidler, J.; McDeavitt, S.; Thompson, M.; Toth, M.; Williamson, M.; Willit, J.

1999-08-12T23:59:59.000Z

219

Final  

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

, , Final for Vegetation Control at VHF Stations, Microwave Stations, Electrical Substations, and Pole Yards . Environmental Assessment Prepared for Southwestern Power Administration U.S. Department of Energy - _ . . . " Prepared by Black & Veatch October 13,1995 ' Table of Contents 1 . 0 Purpose and Need for Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0 Description of the Alternatives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1 Alternative 1 . No Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2 Alternative 2 . Mechanical and Manual Control . . . . . . . . . . . . . . . . . . . 2.3 Alternative 3 . Proposed Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3.1 Foliar Spray Application . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3.2 Soil-Spot Application . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

220

Fiscal Year 2010 Summary Report on the Epsilon-Metal Phase as a Waste Form for 99 Tc  

SciTech Connect (OSTI)

Epsilon metal (?-metal) is generated in nuclear fuel during irradiation. This metal consists of Pd, Ru, Rh, Mo, and some Te. These accumulate at the UO2 grain boundaries as small (ca 5 µm) particles. These metals have limited solubility in the acid used to dissolve fuel during reprocessing and in typical borosilicate glass. These must be treated separately to improve overall waste loading in glass. This low solubility and their survival in 2 Gy-old natural reactors led us to investigate them as a waste form for the immobilization of 99Tc and 107Pd, two very long-lived isotopes.

Strachan, Denis M.; Crum, Jarrod V.; Buck, Edgar C.; Riley, Brian J.; Zumhoff, Mac R.

2010-09-30T23:59:59.000Z

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


221

Reference concepts for the final disposal of LWR spent fuel and other high activity wastes in Spain  

SciTech Connect (OSTI)

Studies over the last three years have been recently concluded with the selection of a reference repository concept for the final disposal of spent fuel and other high activity wastes in deep geological formations. Two non-site specific preliminary designs, at a conceptual level, have been developed; one considers granite as the host rock and the other rock salt formations. The Spanish General Radioactive Waste Program also considers clay as a potential host rock for HLW deep disposal; conceptualization for a deep repository in clay is in the initial phase of development. The salt repository concept contemplates the disposal of the HLW in self-shielding casks emplaced in the drifts of an underground facility, excavated at a depth of 850 m in a bedded salt formation. The Custos Type I(7) cask admits up to seven intact PWR fuel assemblies or 21 of BWR type. The final repository facilities are planned to accept a total of 20,000 fuel assemblies (PWR and BWR) and 50 vitrified waste canisters over a period of 25 years. The total space needed for the surface facilities amounts to 322,000 m{sup 2}, including the rock salt dump. The space required for the underground facilities amounts to 1.2 km{sup 2}, approximately. The granite repository concept contemplates the disposal of the HLW in carbon steel canisters, embedded in a 0.75 m thick buffer of swelling smectite clay, in the drifts of an underground facility, excavated at a depth of 55 m in granite. Each canister can host 3 PWR or 9 BWR fuel assemblies. For this concept the total number of canisters needed amounts to 4,860. The space required for the surface and underground facilities is similar to that of the salt concept. The technical principles and criteria used for the design are discussed, and a description of the repository concept is presented.

Huertas, F.; Ulibarri, A. [ENRESA, Madrid (Spain)

1993-12-31T23:59:59.000Z

222

Forms  

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

are in .pdf format) After Hours Access Policy After Hours Request Form Cleanroom Access Procedures for New Users Deposition Request Form Exit Form Flycutting Request Form Hot...

223

Final Environmental Impact Statement for the Tank Waste Remediation System, Hanford Site, Richland, Washington  

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

for the Tank Waste Remediation System, Hanford Site, Richland, Washington for the Tank Waste Remediation System, Hanford Site, Richland, Washington file:///I|/Data%20Migration%20Task/EIS-0189-FEIS-Summary-1996.HTM[6/27/2011 11:21:59 AM] The National Environmental Policy Act (NEPA) requires Federal agencies to analyze the potential environmental impacts of their proposed actions to assist them in making informed decisions. A similar Washington State law, the State Environmental Policy Act (SEPA), requires State agencies, including the Washington State Department of Ecology (Ecology), to analyze environmental impacts before making decisions that could impact the environment. A major emphasis of both laws is to promote public awareness of these actions and provide opportunities for public involvement. Because NEPA and SEPA requirements are similar, the U.S. Department of Energy (DOE) and Ecology

224

Testing of organic waste surrogate materials in support of the Hanford organic tank program. Final report  

SciTech Connect (OSTI)

To address safety issues regarding effective waste management efforts of underground organic waste storage tanks at the Hanford Site, the Bureau of Mines conducted a series of tests, at the request of the Westinghouse Hanford company. In this battery of tests, the thermal and explosive characteristics of surrogate materials, chosen by Hanford, were determined. The surrogate materials were mixtures of inorganic and organic sodium salts, representing fuels and oxidants. The oxidants were sodium nitrate and sodium nitrite. The fuels were sodium salts of oxalate, citrate and ethylenediamine tetraacetic acid (EDTA). Polyethylene powder was also used as a fuel with the oxidant(s). Sodium aluminate was used as a diluent. In addition, a sample of FeCN, supplied by Hanford was also investigated.

Turner, D.A. [Westinghouse Hanford Co., Richland, WA (United States); Miron, Y. [Bureau of Mines (United States)

1994-01-01T23:59:59.000Z

225

Microbial Gas Generation Under Expected Waste Isolation Pilot Plant Repository Conditions: Final Report  

SciTech Connect (OSTI)

Gas generation from the microbial degradation of the organic constituents of transuranic (TRU) waste under conditions expected in the Waste Isolation Pilot Plant (WIPP) was investigated. The biodegradation of mixed cellulosic materials and electron-beam irradiated plastic and rubber materials (polyethylene, polyvinylchloride, hypalon, leaded hypalon, and neoprene) was examined. We evaluated the effects of environmental variables such as initial atmosphere (air or nitrogen), water content (humid ({approx}70% relative humidity, RH) and brine inundated), and nutrient amendments (nitogen phosphate, yeast extract, and excess nitrate) on microbial gas generation. Total gas production was determined by pressure measurement and carbon dioxide (CO{sub 2}) and methane (CH{sub 4}) were analyzed by gas chromatography; cellulose degradation products in solution were analyzed by high-performance liquid chromatography. Microbial populations in the samples were determined by direct microscopy and molecular analysis. The results of this work are summarized.

Gillow, J.B.; Francis, A.

2011-07-01T23:59:59.000Z

226

Remaining Sites Verification Package for the 100-D-2 Lead Sheeting Waste Site, Waste Site Reclassification Form 2007-030  

SciTech Connect (OSTI)

The 100-D-2 Lead Sheeting waste site was located approximately 50 m southwest of the 185-D Building and approximately 16 m north of the east/west oriented road. The site consisted of a lead sheet covering a concrete pad. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2008-03-19T23:59:59.000Z

227

Studies of waste-canister compatibility. [Waste forms: Al-Si and Pb-Sn matrix alloys, FUETAP, glass, Synroc D, and waste particles coated with carbon or carbon plus SiC  

SciTech Connect (OSTI)

Compatibility studies were conducted between 7 waste forms and 15 potential canister structural materials. The waste forms were Al-Si and Pb-Sn matrix alloys, FUETAP, glass, Synroc D, and waste particles coated with carbon or carbon plus silicon carbide. The canister materials included carbon steel (bare and with chromium or nickel coatings), copper, Monel, Cu-35% Ni, titanium (grades 2 and 12), several Inconels, aluminum alloy 5052, and two stainless steels. Tests of either 6888 or 8821 h were conducted at 100 and 300/sup 0/C, which bracket the low and high limits expected during storage. Glass and FUETAP evolved sulfur, which reacted preferentially with copper, nickel, and alloys of these metals. The Pb-Sn matrix alloy stuck to all samples and the carbon-coated particles to most samples at 300/sup 0/C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 800/sup 0/C was included because it is representative of a transportation accident and is required of casks containing nuclear materials. During these tests (1) glass and FUETAP evolved sulfur, (2) FUETAP evolved large amounts of gas, (3) Synroc stuck to titanium alloys, (4) glass was molten, and (5) both matrix alloys were molten with considerable chemical interactions with many of the canister samples. If this test condition were imposed on waste canisters, it would be design limiting in many waste storage concepts.

McCoy, H.E.

1983-01-01T23:59:59.000Z

228

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

SciTech Connect (OSTI)

The 120-F-1 waste site consisted of two dumping areas located 660 m southeast of the 105-F Reactor containing laboratory equipment and bottles, demolition debris, light bulbs and tubes, small batteries, small drums, and pesticide contaminated soil. It is probable that 108-F was the source of the debris but the material may have come from other locations within the 100-F Area. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

J. M. Capron

2008-06-27T23:59:59.000Z

229

New data on mineral forms of rare metals in phosphogypsum wastes  

Science Journals Connector (OSTI)

Phosphogypsum is an industrial waste of the processing ... . This is a valuable and promising technogenous rare-metal feedstock. The samples of fresh and old phosphogypsum were studied using precision physical te...

A. E. Samonov

2011-09-01T23:59:59.000Z

230

Spent nuclear fuel as a waste form for geologic disposal: Assessment and recommendations on data and modeling needs  

SciTech Connect (OSTI)

This study assesses the status of knowledge pertinent to evaluating the behavior of spent nuclear fuel as a waste form in geologic disposal systems and provides background information that can be used by the DOE to address the information needs that pertain to compliance with applicable standards and regulations. To achieve this objective, applicable federal regulations were reviewed, expected disposal environments were described, the status of spent-fuel modeling was summarized, and information regarding the characteristics and behavior of spent fuel was compiled. This compiled information was then evaluated from a performance modeling perspective to identify further information needs. A number of recommendations were made concerning information still needed to enhance understanding of spent-fuel behavior as a waste form in geologic repositories. 335 refs., 22 figs., 44 tabs.

Van Luik, A.E.; Apted, M.J.; Bailey, W.J.; Haberman, J.H.; Shade, J.S.; Guenther, R.E.; Serne, R.J.; Gilbert, E.R.; Peters, R.; Williford, R.E.

1987-09-01T23:59:59.000Z

231

Systematic Process Synthesis and Design Methods for Cost Effective Waste Minimization. Final report  

SciTech Connect (OSTI)

This report focuses on research done over the past four years under the grant with the above title. In addition, the report also includes a brief summary of work done before 1994 under grant DOE-DE-FG02-85ER13396. Finally, a complete list of publications that acknowledge support from this grant is listed at the end.

Biegler, L.T.; Grossmann, I.E.; Westerberg, A.W.

1998-02-14T23:59:59.000Z

232

GEOHYDROLOGICAL STUDIES FOR NUCLEAR WASTE ISOLATION AT THE HANFORD RESERVATION -- Vol. I: Executive Summary; Vol. II: Final Report  

E-Print Network [OSTI]

long-term feasibility study conducted by Rockwell's Basalt Waste Isolation Program for the National Waste Terminal Storagelong-term feasibility study conducted by Rockwell's Basalt Waste Isolation Program for the National Was~e Terminal Storage

Apps, J.

2010-01-01T23:59:59.000Z

233

FINAL ENVIRONMENTAL ASSESSMENT FOR REMOVAL ACTIONS AT THE TECHNICAL AREA III CLASSIFIED WASTE LANDFILL, SANDIA NATIONAL LABORATORIES, NEW MEXICO - DOE/EA-1729  

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

FINAL ENVIRONMENTAL ASSESSMENT FOR REMOVAL FINAL ENVIRONMENTAL ASSESSMENT FOR REMOVAL ACTIONS AT THE TECHNICAL AREA III CLASSIFIED WASTE LANDFILL, SANDIA NATIONAL LABORATORIES, NEW MEXICO DOE/EA-1729 August 2010 National Nuclear Security Administration Sandia Site Office P.O. Box 5400 Albuquerque, New Mexico 87185-5400 DOE/EA-1729: Environmental Assessment for Removal Actions at the Technical Area III August 2010 Classified Waste Landfill, Sandia National Laboratories, New Mexico i TABLE OF CONTENTS Section 1.0 PURPOSE AND NEED FOR AGENCY ACTION .................................................................... Page 1 1.1 Background .................................................................................................................................. 1

234

EIS-0203F; DOE Programmatic Spent Nuclear Fuel Management and INEL Environmental Restoration and Waste Management Programs Final Environmental Impact Statement  

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

Summary-1995.html[6/27/2011 12:08:32 PM] Summary-1995.html[6/27/2011 12:08:32 PM] SUMMARY DOE/EIS-0203-F Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement Summary April 1995 U.S. Department of Energy Office of Environmental Management Idaho Operations Office Department of Energy Washington, DC 20585 April 1995 Dear Citizen: This is a summary of the Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement. The Department of Energy and

235

Technical justifications for the tests and criteria in the waste form technical position appendix on cement stabilization  

SciTech Connect (OSTI)

As part of its technical assistance to the Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) developed a background document for the cement stabilization appendix, Appendix A, to Rev. 1 of the Technical Position on Waste Form (TP). Here we present an overview of this background document, which provides technical justification for the stability tests to be performed on cement-stabilized waste forms and for the criteria posed in each test, especially for those tests which have been changed from their counterparts in the May 1983 Rev. 0 TP. We address guidelines for procedures from Appendix A which are considered in less detail or not at all in the Rev. 0 of the TP, namely, qualification specimen preparation (mixing, curing, storage), statistical sampling and analysis, process control program specimen preparation and examination, and surveillance specimens. For each waste form qualification test, criterion or procedural guidelines, we consider the reason for its inclusion in Appendix A, the changes from Rev. 0 of the TP (if applicable), and a discussion of the justification or rationale for these changes.

Siskind, B.; Cowgill, M.G.

1992-04-01T23:59:59.000Z

236

Technical justifications for the tests and criteria in the waste form technical position appendix on cement stabilization  

SciTech Connect (OSTI)

As part of its technical assistance to the Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) developed a background document for the cement stabilization appendix, Appendix A, to Rev. 1 of the Technical Position on Waste Form (TP). Here we present an overview of this background document, which provides technical justification for the stability tests to be performed on cement-stabilized waste forms and for the criteria posed in each test, especially for those tests which have been changed from their counterparts in the May 1983 Rev. 0 TP. We address guidelines for procedures from Appendix A which are considered in less detail or not at all in the Rev. 0 of the TP, namely, qualification specimen preparation (mixing, curing, storage), statistical sampling and analysis, process control program specimen preparation and examination, and surveillance specimens. For each waste form qualification test, criterion or procedural guidelines, we consider the reason for its inclusion in Appendix A, the changes from Rev. 0 of the TP (if applicable), and a discussion of the justification or rationale for these changes.

Siskind, B.; Cowgill, M.G.

1992-01-01T23:59:59.000Z

237

GEOHYDROLOGICAL STUDIES FOR NUCLEAR WASTE ISOLATION AT THE HANFORD RESERVATION -- Vol. I: Executive Summary; Vol. II: Final Report  

E-Print Network [OSTI]

NUCLEAR WASTE ISOLATION AT THE HANFORD RESERVATION Volume I:of Washington state." Rockwell Hanford Operations Topicalmodel evaluation at the Hanford nuclear waste facility."

Apps, J.

2010-01-01T23:59:59.000Z

238

Environmental Solutions, A Summary of Contributions for CY04: Battelle Contributions to the Waste Treatment Plant  

SciTech Connect (OSTI)

In support of the Waste Treatment Plant (WTP), Battelle conducted tests on mixing specific wastes within the plant, removing troublesome materials from the waste before treatment, and determining if the final waste forms met the established criteria. In addition, several Battelle experts filled full-time positions in WTP's Research and Testing and Process and Operations departments.

Beeman, Gordon H.

2005-03-08T23:59:59.000Z

239

Hydrogen generation by metal corrosion in simulated Waste Isolation Pilot Plant environments. Final report  

SciTech Connect (OSTI)

The corrosion and gas-generation characteristics of four material types: low-carbon steel (the current waste packaging material for the Waste Isolation Pilot Plant), Cu-base and Ti-base (alternative packaging) materials, and Al-base (simulated waste) materials were determined in both the liquid and vapor phase of Brine A, a brine representative of an intergranular Salado Formation brine. Test environments consisted primarily of anoxic brine with overpressures of N{sub 2}, CO{sub 2}, H{sub 2}S, and H{sub 2}. Limited tests of low-carbon steel were also performed in simulated-backfill environments and in brine environments with pH values ranging from 3 to 11. Low-carbon steel reacted at a slow, measurable rate with anoxic brine, liberating H{sub 2} on an equimolar basis with Fe reacted. Presence of CO{sub 2} caused the initial reaction to proceed more rapidly, but CO{sub 2}-induced passivation stopped the reaction if the CO{sub 2} were present in sufficient quantities. Addition of H{sub 2}S to a CO{sub 2}-passivated system caused reversal of the passivation. Low-carbon steel immersed in brine with H{sub 2}S showed no reaction, apparently because of passivation of the steel by formation of FeS. Addition of CO{sub 2} to an H{sub 2}S-passivated system did not reverse the passivation. Cu- and Ti-base materials showed essentially no corrosion when exposed to brine and overpressures of N{sub 2}, CO{sub 2}, and H{sub 2}S except for the rapid and complete reaction between Cu-base materials and H{sub 2}S. The Al-base materials reacted at approximately the same rate as low-carbon steel when immersed in anoxic Brine A; considerably more rapidly in the presence of CO{sub 2} or H{sub 2}S; and much more rapidly when iron was present in the system as a brine contaminant. High-purity Al was much more susceptible to corrosion than the 6061 alloy. No significant reaction took place on any material in any environment in the vapor-phase exposures.

Telander, M.R.; Westerman, R.E. [Battelle Pacific Northwest Lab., Richland, WA (United States)

1997-03-01T23:59:59.000Z

240

Final Environmental Impact Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina  

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

Impact Impact \ DoE/Els-o120 Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina Volume 3 Q ~+ ~ FNT O&@+@ &v a w ~ k ~ ;%." $ +6 & ~+e $TiTES Of December 1987 United States Department of Energy TABLE OF CONTENTS ~pendix G ASSESSMENT OF ALTERNATIVE STRATEGIES FOR STORAGE FACILITIES . . . . . . . . . . . G.1 No-Action Strategy . . . . . . . . G.1.l Sununarv and Objectives . . NEW DISPOSAL/ . . . . . . . . . . . . . . . G.1.2 Groundwater and Surface Water Effects G.1.3 Nonradioactive Atmospheric Releases . G.1.4 Ecological Effects . . . . . . . . . G.1.5 Radiological Releases . . . . . . . . G.1.6 Archaeological and Historic Resources G.1.7 SOciOecOnOmics . . . . . . . . . . . G.1.8 Dedication of Site . . . . . . . . . G.1.9 Institutional Impacts . . . . . . . . G.l.10 Noise . . . . . . . . . . . . . . . . G.2 Dedication Strategy . . . . . . . . . . . . . G.2.1 G.2.2 G.2.3 G.2.4 G.2.5 G.2.6

Note: This page contains sample records for the topic "final waste form" 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

Portable sensor for hazardous waste. Final report, March 31, 1995--May 31, 1997  

SciTech Connect (OSTI)

This report summarizes accomplishments for the second phase of a 5-year program designed to develop a portable monitor for sensitive hazardous waste detection. The approach is to excite atomic fluorescence by the technique of Spark-Induced Breakdown Spectroscopy (SIBS). The principal goals for this second phase of the program were to demonstrate sensitive detection of additional species, both RCRA metals (Sb, Be, Cd, Cr, Pb, As, Hg) and radionuclides (U, Th, Tc); to identify potential applications and develop instrument component processes, including, sample collection and excitation, measurement and test procedures, and calibration procedures; and to design a prototype instrument. Successful completion of these task results in being able to fabricate and field test a prototype of the instrument during the program`s third phase.

Piper, L.G.; Hunter, A.J.R.; Fraser, M.E.; Davis, S.H.; Finson, M.L.

1997-12-31T23:59:59.000Z

242

Flowsheet model for the electrochemical treatment of liquid radioactive wastes. Final report  

SciTech Connect (OSTI)

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

Hobbs, D.T. [Westinghouse Savannah River Co., Aiken, SC (United States); Prasad, S.; Farell, A.E.; Weidner, J.W.; White, R.E. [South Carolina Univ., Columbia, SC (United States). Dept. of Chemical Engineering

1995-12-31T23:59:59.000Z

243

Savannah River Site High-Level Waste Tank Closure, Final Environmental Impact Statement  

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

TANK FARM DESCRIPTION AND CLOSURE PROCESS TANK FARM DESCRIPTION AND CLOSURE PROCESS DOE/EIS-0303 Tank Farm Description FINAL May 2002 and Closure Process A-iii TABLE OF CONTENTS Section Page A.1 Introduction........................................................................................................................... A-1 A.2 Overview of SRS HLW Management .................................................................................. A-1 A.3 Description of the Tank Farms ............................................................................................. A-4 A.3.1 Tanks........................................................................................................................ A-4 A.3.2 Evaporator Systems .................................................................................................

244

Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (DOE/EIS-0287) (11/28/06)  

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

811 Federal Register 811 Federal Register / Vol. 71, No. 228 / Tuesday, November 28, 2006 / Notices Information Relay Service (FIRS) at 1-800-877-8339. [FR Doc. E6-20124 Filed 11-27-06; 8:45 am] BILLING CODE 4000-01-P DEPARTMENT OF ENERGY Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement AGENCY: Department of Energy. ACTION: Amended Record of Decision. SUMMARY: The U.S. Department of Energy (DOE) is amending its Record of Decision (ROD) published December 19, 2005 (70 Federal Register [FR] 75165), pursuant to the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (Final EIS) (DOE/EIS-0287, September 2002). The Final EIS analyzed two sets of alternatives for accomplishing DOE's

245

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

SciTech Connect (OSTI)

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

Jasen, W.G.

1998-01-07T23:59:59.000Z

246

Evaluation of standard durability tests towards the qualification process for the glass-zeolite ceramic waste form  

SciTech Connect (OSTI)

Glass-bonded zeolite is being developed as a potential ceramic waste form for the disposition of radionuclides associated with the Department of Energy`s (DOE`s) spent nuclear fuel conditioning activities. The utility of several standard durability tests was evaluated as a first step in developing methods and criteria that can be applied towards the process of qualifying this material for acceptance into the DOE Civilian Radioactive Waste Management System. The effects of pH, leachant composition, and sample surface-area-to leachant-volume ratios on the durability test results are discussed, in an attempt to investigate the release mechanisms and other physical and chemical parameters that are important for the acceptance criteria, including the establishment of appropriate test methodologies required for product consistency measurements.

Simpson, L.J.; Wronkiewicz, D.J. [Chemical Technology Division, Argonne National Laboratory (Illinois)

1996-12-31T23:59:59.000Z

247

Microsoft Word - Oneida Waste to Energy Project DOE Final EA 1862  

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

ONEIDA SEVEN GENERATIONS ONEIDA SEVEN GENERATIONS CORPORATION: ENERGY RECOVERY PROJECT, GREEN BAY, WISCONSIN U.S. Department of Energy Office of Energy Efficiency and Renewable Energy Golden Field Office NOVEMBER 2011 DOE/EA-1862 FINAL ENVIRONMENTAL ASSESSMENT FOR THE ONEIDA SEVEN GENERATIONS CORPORATION: ENERGY RECOVERY PROJECT, GREEN BAY, WISCONSIN U.S. Department of Energy Office of Energy Efficiency and Renewable Energy Golden Field Office NOVEMBER 2011 DOE/EA-1862 DOE/EA-1862 iii November 2011 COVER SHEET RESPONSIBLE AGENCY: U.S. Department of Energy TITLE: Environmental Assessment for Oneida Seven Generations Corporation: Energy Recovery Project, Green Bay, Wisconsin (DOE/EA-1862) CONTACT: For more information on this Environmental Assessment (EA), please contact:

248

Idaho High-Level Waste & Facilities Disposition, Final Environmental Impact Statement  

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

Environmental Impact Statement: A detailed environmental analysis for any proposed major Federal action that could significantly affect the quality of the human environment. A tool to assist in decision-making, it describes the positive and negative environmental effects of the proposed undertaking and alternatives. A draft EIS is issued, followed by a final EIS. Scoping: An early and open process in which the public is invited to participate in identi- fying issues and alternatives to be con- sidered in this EIS. DOE allows a minimum of 30 days for the receipt of public comments. Alternatives: A range of courses of action that would meet the agency's purpose and need for action. NEPA requires that an EIS con- sider a No Action Alternative. Comment Period: A regulatory minimum 45-day

249

Coolside Waste Management Research. Final report, April 23, 1991--June 30, 1996  

SciTech Connect (OSTI)

This study was initiated during a successful test of the Coolside flue gas desulfurization technology at Ohio Edison`s Edgewater generating station in 1991. Coolside is a lime duct technology which is installed on the downstream side of the last heat exchanger. As tested by Ohio Edison, it also employs an alkali reagent, in this case NaOH, to enhance sulfur capture. The overall goal of this study was to develop sufficient chemical and physical data to insure the environmentally safe disposal of the material. This final report summarizes the important aspects of the project, but it does not present all of the data that was produced. Further details may be found in the monthly and quarterly reports that were filed with the Morgantown Energy Technology Center. This report is organized into six chapters which present the important conclusions of the principal areas of investigation.

NONE

1998-12-31T23:59:59.000Z

250

Final Meeting Summary Page 1  

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

... 4 Final Tank Closure and Waste Management Environmental Impact Statement (TC & WM EIS) (joint with PIC) ......

251

Surface Roughening in Formed Al Sheets for Automative Applications. Final Report  

SciTech Connect (OSTI)

New experimental and analytical tools for characterizing various types of forming-induced surface roughening have been developed and used to study the evolution of surface roughening in 6000 series aluminum sheets in the T4 condition. Particular attention was focused on the evolution of the ''diamond defect,'' which often requires a subsequent mechanical surface treatment to achieve an acceptable surface finish. Testing was primarily done in plane strain tension and samples evaluated at various strain levels up to and including the forming limit. Characterization techniques included optical and scanning electron microscopy, white light phase shift interferometry, and orientation imaging microscopy. Roughening patterns were observed to communicate through the sheet thickness, e.g., peaks on one side correspond to peaks on the other; valleys to valleys. At the grain or grain cluster scale, roughening was found to be governed by the Schmid rather than the Taylor factor.

Piehler, Henry R.

2003-07-23T23:59:59.000Z

252

Standard practice for prediction of the long-term behavior of materials, including waste forms, used in engineered barrier systems (EBS) for geological disposal of high-level radioactive waste  

E-Print Network [OSTI]

1.1 This practice describes test methods and data analyses used to develop models for the prediction of the long-term behavior of materials, such as engineered barrier system (EBS) materials and waste forms, used in the geologic disposal of spent nuclear fuel (SNF) and other high-level nuclear waste in a geologic repository. The alteration behavior of waste form and EBS materials is important because it affects the retention of radionuclides by the disposal system. The waste form and EBS materials provide a barrier to release either directly (as in the case of waste forms in which the radionuclides are initially immobilized), or indirectly (as in the case of containment materials that restrict the ingress of groundwater or the egress of radionuclides that are released as the waste forms and EBS materials degrade). 1.1.1 Steps involved in making such predictions include problem definition, testing, modeling, and model confirmation. 1.1.2 The predictions are based on models derived from theoretical considerat...

American Society for Testing and Materials. Philadelphia

2007-01-01T23:59:59.000Z

253

DOE/EIS-0287-SA-01: Supplement Analysis for the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (June 2005)  

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

7 7 -SA-Ol SUPPLEMENT ANALYSIS For The Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement June 2005 United States Department of Energy Idaho Operations Office 1.0 2.0 3.0 4.0 5.0 6.0 DOEÆIS-0287 -SA-O 1 TABLE OF CONTENTS Introduction......................................................................................................................... 4 Background......................................................................................................................... 4 Areas of Review.................................................................................................................. 6 3.1 3.2 3.3 3.4 Proposed Waste Treatment Technology.......... .......................................................

254

Thermal processing systems for TRU mixed waste  

SciTech Connect (OSTI)

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

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

1992-01-01T23:59:59.000Z

255

Thermal processing systems for TRU mixed waste  

SciTech Connect (OSTI)

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

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

1992-08-01T23:59:59.000Z

256

Emergence of interest groups on hazardous waste siting: how do they form and survive  

SciTech Connect (OSTI)

This paper discusses the two components of the facilitative setting that are important for group formation. The first component, the ideological component, provides the basic ideas that are adopted by the emerging group. The ideological setting for group formation is produced by such things as antinuclear news coverage and concentration of news stories on hazardous waste problems, on ideas concerning the credibility of the federal government, and on the pervasivensee of ideas about general environmental problems. The organizational component of the facilitative setting provides such things as leadership ability, flexible time, resources, and experience. These are important for providing people, organization, and money to achieve group goals. By and large, the conditions conducive to group formation, growth, and survival are outside the control of decision-makers. Agencies and project sponsors are currently caught in a paradox. Actively involving the public in the decision-making process tends to contribute to the growth and survival of various interest groups. Not involving the public means damage to credibility and conflict with values concerning participatory democracy. Resolution in this area can only be achieved when a comprehensive, coordinated national approach to hazardous waste management emerges. 26 refs.

Williams, R.G.; Payne, B.A.

1985-10-30T23:59:59.000Z

257

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

SciTech Connect (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

258

EIS-0203F; DOE Programmatic Spent Nuclear Fuel Management and INEL Environmental Restoration and Waste Management Programs Final Environmental Impact Statement  

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

02-1995/voli.html[6/27/2011 12:23:34 PM] 02-1995/voli.html[6/27/2011 12:23:34 PM] DOE Programmatic Spent Nuclear Fuel Management and INEL Environmental Restoration and Waste Management Programs Final Environmental Impact Statement VOLUME II VOLUME II Part A COVER SHEET RESPONSIBLE AGENCIES: Lead Federal Agency: U.S. Department of Energy Cooperating Federal Agency: U.S. Department of the Navy TITLE: Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement. CONTACT: For further information on this Environmental Impact Statement call or contact: DOE Idaho Operations Office Bradley P. Bugger Office of Communications 850 Energy Drive, MS 1214 Idaho Falls, ID 83403-3189

259

Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada  

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

Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada DOE/EIS-0250 Errata Sheet Since release of the Final EIS for Yucca Mountain on February 14, 2002 as part of the Site Recommendation documentation required under the Nuclear Waste Policy Act, as amended, the Department of Energy (DOE) has identified a variety of errors in the document. These errors were found to include: editing errors - errors in editorial style, rounding, and unit conversions data entry errors, errors in typing a number transcription errors - errors in transcribing information from one part of the document to another, failures to update the text from the most current analyses at the time of the

260

EIS-0203F; DOE Programmatic Spent Nuclear Fuel Management and INEL Environmental Restoration and Waste Management Programs Final Environmental Impact Statement  

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

02-1995/voli.html[6/27/2011 12:23:34 PM] 02-1995/voli.html[6/27/2011 12:23:34 PM] DOE Programmatic Spent Nuclear Fuel Management and INEL Environmental Restoration and Waste Management Programs Final Environmental Impact Statement VOLUME II VOLUME II Part A COVER SHEET RESPONSIBLE AGENCIES: Lead Federal Agency: U.S. Department of Energy Cooperating Federal Agency: U.S. Department of the Navy TITLE: Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement. CONTACT: For further information on this Environmental Impact Statement call or contact: DOE Idaho Operations Office Bradley P. Bugger Office of Communications 850 Energy Drive, MS 1214 Idaho Falls, ID 83403-3189

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


261

Fundamental properties of monolithic bentonite buffer material formed by cold isostatic pressing for high-level radioactive waste repository  

SciTech Connect (OSTI)

The methods of fabrication, handling, and emplacement of engineered barriers used in a deep geological repository for high level radioactive waste should be planned as simply as possible from the engineering and economic viewpoints. Therefore, a new concept of a monolithic buffer material around a waste package have been proposed instead of the conventional concept with the use of small blocks, which would decrease the cost for buffer material. The monolithic buffer material is composed of two parts of highly compacted bentonite, a cup type body and a cover. As the forming method of the monolithic buffer material, compaction by the cold isostatic pressing process (CIP) has been employed. In this study, monolithic bentonite bodies with the diameter of about 333 mm and the height of about 455 mm (corresponding to the approx. 1/5 scale for the Japanese reference concept) were made by the CIP of bentonite powder. The dry densities: {rho}d of the bodies as a whole were measured and the small samples were cut from several locations to investigate the density distribution. The swelling pressure and hydraulic conductivity as function of the monolithic body density for CIP-formed specimens were also measured. High density ({rho}d: 1.4--2.0 Mg/m{sup 3}) and homogeneous monolithic bodies were formed by the CIP. The measured results of the swelling pressure (3--15 MPa) and hydraulic conductivity (0.5--1.4 x 10{sup {minus}13} m/s) of the specimens were almost the same as those for the uniaxial compacted bentonite in the literature. It is shown that the vacuum hoist system is an applicable handling method for emplacement of the monolithic bentonite.

Kawakami, S.; Yamanaka, Y.; Kato, K.; Asano, H.; Ueda, H.

1999-07-01T23:59:59.000Z

262

Long-term, low-level radwaste volume-reduction strategies. Volume 4. Waste disposal costs. Final report  

SciTech Connect (OSTI)

Volume 4 establishes pricing levels at new shallow land burial grounds. The following conclusions can be drawn from the analyses described in the preceding chapters: Application of volume reduction techniques by utilities can have a significant impact on the volumes of wastes going to low-level radioactive waste disposal sites. Using the relative waste stream volumes in NRC81 and the maximum volume reduction ratios provided by Burns and Roe, Inc., it was calculated that if all utilities use maximum volum

Sutherland, A.A.; Adam, J.A.; Rogers, V.C.; Merrell, G.B.

1984-11-01T23:59:59.000Z

263

Final Report Waste Incineration  

E-Print Network [OSTI]

regulations and the formation of dioxins, as well as a big effort is put into the treatment of solid residues

264

Application of microwave solidification technology to radioactive waste  

SciTech Connect (OSTI)

The EPA has declared vitrification to be the Best Available Demonstrated Technology (BDAT) for High Level Radioactive Waste (40 CFR 268.42). Vitrification has been chosen as the method of choice for treating a number of radioactive residues and wastes in the DOE complex. Vitrification offers advantages of waste volume reduction, the ability to handle changing waste forms, and a stable, nonleachable final waste form. Microwave heating is a superior method for vitrification of radioactive wastes. Advantages of microwave heating include: (1) direct waste heating, eliminates need for electrodes, refractories and other consumables; (2) ``in-can`` processing allows for treatment of the material in its final container, (3) a mechanically simple system where the microwaves are generated away from the treatment area and transmitted to the treatment applicator by a wave guide, thus minimizing worker exposure to radiation; (4) easier equipment maintenance; and (5) a high degree of public acceptance.

Harris, M.; Sprenger, G.; Roushey, B.; Fenner, G.; Nieweg, R.

1995-09-28T23:59:59.000Z

265

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

SciTech Connect (OSTI)

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

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

2012-04-01T23:59:59.000Z

266

Superlattice Structure and Precipitates in O+ and Zr+ Ion Coimplanted SrTiO3: a Model Waste Form for 90Sr  

SciTech Connect (OSTI)

We investigate strontium titanate as a model waste form for 90Sr. Implantation with O+ and Zr+ ions, followed by annealing at 1423 K, was performed to simulate 90Sr to 90Zr decays. At low Zr concentrations, we observe formation of a ZrO-Sr superlattice structure. Ab initio calculations indicate that this atomic configuration is energetically favorable. At higher Zr concentrations, we observe precipitates of ZrO2 with a coherently strained interface, or a monolayer of disordered interfacial structure. Potential candidacy of 90SrTiO3 as a waste form for permanent disposal of 90Sr is discussed.

Jiang, Weilin; Van Ginhoven, Renee M.; Kovarik, Libor; Jaffe, John E.; Arey, Bruce W.

2012-07-13T23:59:59.000Z

267

Assessment of ocean waste disposal. Task 5. Human-health impacts of waste constituents. 2. Pathogens and antibiotic- and heavy-metal-resistant bacteria. Final report  

SciTech Connect (OSTI)

Disposal of wastes in the ocean has been practiced by coastal nations for many decades. All areas of the ocean have been subject to disposal use, including estuaries, nearshore, open shelf, and deep ocean sites. Until recently, it was believed that pathogenic bacteria did not survive for any significant period of time in estuarine and marine environments. Scientists and public-health workers never bothered to ask the question could viable, virulent pathogens be present in water samples even though they could not be detected by conventional plating methods. This laboratory answered this question in the affirmative for several bacterial pathogens, and this is discussed in detail. What follows in the report is a description of potentially harmful constituents of wastes, ways in which those constituents could reach humans, known incidents of human disease contracted from wastes, detection of waste-borne disease agents, management technologies, and monitoring and predictive technologies. Since the report is not just a review of the literature, not all known literature has been discussed. However, every attempt is made to include very relevant material, regardless of its age. What follows then is both a literature review and a position paper.

Grimes, D.J.

1986-05-01T23:59:59.000Z

268

EIS-0287: Final Environmental Impact Statement | Department of Energy  

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

Final Environmental Impact Statement Final Environmental Impact Statement EIS-0287: Final Environmental Impact Statement Idaho High-Level Waste and Facilities Disposition This EIS analyzes the potential environmental consequences of alternatives for managing high- level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liq- uid waste at the Idaho National Engineering and Environmental Laboratory (INEEL) in liquid and solid forms. This EIS also analyzes alternatives for the final disposition of HLW management facilities at the INEEL after their missions are completed. After considering comments on the Draft EIS (DOE/EIS- 0287D), as well as information on available treatment technologies, DOE and the State of Idaho have iden- tified separate preferred alternatives for

269

Preliminary characterization of deposits formed on super heater surfaces in an FBC-boiler fired with municipal solid waste  

SciTech Connect (OSTI)

A preliminary study of the chemical and mineralogical composition of deposits formed on super heater tubes in a CFB fired with 100% sorted municipal solid waste has been carried out. Samples of deposits formed on both the windward and leeward side of the tubes were analyzed with the aim to identify the ash species involved in fouling and to get information about chemical interaction between the tube alloys and the deposits. The metal temperatures in the super heater region were in the range 460--540 C during the sampling period. The identified deposit constituents show the importance of alkali metal chlorides in the deposit forming process. Alkali metal chlorides (NaCl and KCl) were found both on the windward side deposits and on the leeward side. Other components were CaSO{sub 4}m MgO and some oxide and phosphate compounds. Some of these components have probably been formed through reaction between the alloy and the deposit but more work will be done in co-operation with the Competence Centre for High Temperature Corrosion, Sweden in order to elucidate such interactions and the influence of deposits on the corrosion rates. The presence of chlorides on an alloyed steel at the temperatures used here may cause a rapid deterioration of the protective oxide scale on the alloy. First, a layer of molten chlorides may dissolve species from the protective oxide layer on the steel tube. Secondly, corrosion may occur according to a mechanism called active oxidation, which involves diffusion of chlorine to the metal/oxide interface and breakdown of the scale due to formation of new products.

Steenari, B.M.; Lindqvist, O.; Andersson, B.A.

1999-07-01T23:59:59.000Z

270

National Low-Level Waste Management Program final summary report of key activities and accomplishments for fiscal year 1997  

SciTech Connect (OSTI)

The US Department of Energy (DOE) has responsibilities under the Low-Level Radioactive Waste Policy Amendments Act of 1985 to assist states and compacts in their siting and licensing efforts for low-level radioactive waste disposal facilities. The National Low-Level Waste Management Program (NLLWMP) is the element of the DOE that performs the key support activities under the Act. The NLLWMP`s activities are driven by the needs of the states and compacts as they prepare to manage their low-level waste under the Act. Other work is added during the fiscal year as necessary to accommodate new requests brought on by status changes in states` and compacts` siting and licensing efforts. This report summarizes the activities and accomplishments of the NLLWMP during FY 1997.

Rittenberg, R.B.

1998-03-01T23:59:59.000Z

271

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

SciTech Connect (OSTI)

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

Ebert, W. E.

2006-01-31T23:59:59.000Z

272

National Low-Level Waste Management Program final summary report of key activities and accomplishments for fiscal year 1995  

SciTech Connect (OSTI)

To assist the Department of Energy (DOE) in fulfilling its responsibilities under the Low-Level Radioactive Waste Policy Amendments Act of 1985, the National Low-Level Waste Management Program (NLLWMP) outlines the key activities that the NLLWMP will accomplish in the following fiscal year. Additional activities are added during the fiscal year as necessary to accomplish programmatic goals. This report summarizes the activities and accomplishments of the NLLWMP during fiscal year 1995.

Forman, S.

1995-12-01T23:59:59.000Z

273

National Low-Level Waste Management Program final summary report of key activities and accomplishments for fiscal year 1996  

SciTech Connect (OSTI)

To assist the Department of Energy (DOE) in fulfilling its responsibilities under the Low-Level Radioactive Waste Policy Amendments Act of 1985, the National Low-Level Waste Management Program (NLLWMP) outlines the key activities tat the NLLWMP will accomplish in the following fiscal year. Additional activities are added during the fiscal year as necessary to accomplish programmatic goals. This report summarizes the activities and accomplishments of the NLLWMP during Fiscal Year 1996.

Garcia, R.S.

1996-12-01T23:59:59.000Z

274

Final environmental impact statement. Management of commercially generated radioactive waste. Volume 3. Public comments hearing board report  

SciTech Connect (OSTI)

This EIS analyzes the significant environmental impacts that could occur if various technologies for management and disposal of high-level and transuranic wastes from commercial nuclear power reactors were to be developed and implemented. This EIS will serve as the environmental input for the decision on which technology, or technologies, will be emphasized in further research and development activities in the commercial waste management program. The action proposed in this EIS is to (1) adopt a national strategy to develop mined geologic repositories for disposal of commercially generated high-level and transuranic radioactive waste (while continuing to examine subseabed and very deep hole disposal as potential backup technologies) and (2) conduct a R and D program to develop such facilities and the necessary technology to ensure the safe long-term containment and isolation of these wastes. The Department has considered in this statement: development of conventionally mined deep geologic repositories for disposal of spent fuel from nuclear power reactors and/or radioactive fuel reprocessing wastes; balanced development of several alternative disposal methods; and no waste disposal action. This volume contains written public comments and hearing board responses and reports offered on the draft statement.

Not Available

1980-10-01T23:59:59.000Z

275

SRS - Programs - Liquid Waste Disposition  

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

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

276

Yucca Mountain project : FY 2006 annual report for waste form testingactivities.  

SciTech Connect (OSTI)

This report describes the experimental work performed at Argonne National Laboratory (Argonne) during fiscal year (FY 2006) under the Bechtel SAIC Company, LLC (BSC) Memorandum Purchase Order (MPO) contract number B004210CM3X. Because this experimental work is focused on the dissolution and precipitation behavior of neptunium, the report also includes, or incorporates by reference, earlier results that are relevant to presenting a more complete understanding of the likely behavior of neptunium under experimental conditions relevant to the Yucca Mountain repository. Important results relevant to the technical bases, validations, and conservatisms in current source term models are summarized. The CSNF samples were observed to corrode following the general contour of the surface rather than via (for instance) grain boundary attack. This supports the current approach of estimating the effective surface area of corroding CSNF based on the geometric surface area of fuel pellet fragments. It was observed that the neptunium and plutonium concentrations in corroded CSNF samples were somewhat higher at and near the corrosion front (i.e., at the interface between the alteration product ''rind'' layer and the underlying fuel) than in the bulk fuel. The neptunium and plutonium at the corrosion front and in the uranyl alteration layer were found to be in the quadravalent (4+) oxidation state. The uranyl phases that constitute most of the alteration rind were depleted in neptunium relative to the bulk fuel: neptunium concentrations in the uranyl alteration rind were less than 20% of that in the parent fuel. Homogeneous precipitation tests have shown that solids precipitate from a 1 x 10{sup -4} M Np(V) solution over the temperature range of 200-280 C, but no evidence was found that any solids precipitated from the same solution at 150 C through 289 days. The solids formed in the homogeneous precipitation tests were predominantly a Np(IV)-bearing phase, probably NpO{sub 2}. The presence of UO{sub 2} resulted in the rapid precipitation at room temperature of similar amounts of Np(IV)- and Np(V)-bearing phases, probably NpO{sub 2} and Np{sub 2}O{sub 5}. Although the UO{sub 2} is presumed to act as a reducing agent for Np(V) that leads to the precipitation of a Np(IV)-bearing phase, the observed formation of a Np(V)-bearing phase suggests that the UO{sub 2} also catalyzes Np{sub 2}O5 precipitation under these test conditions.

Ebert, W. L.; Fortner, J. A.; Guelis, A. V.; Cunnane, J. C.

2006-11-01T23:59:59.000Z

277

Relevance of biotic pathways to the long-term regulation of nuclear waste disposal: Phase 2, Final report  

SciTech Connect (OSTI)

The results reported here establish the relevance and propose a method for including biotic transport in the assessment and licensing process for commercial low-level waste disposal sites. Earlier work identified the biotic transport mechanisms and process scenarios linking biotic transport with dose to man, and developed models for assessment of impacts. Model modification and improvement efforts in enhancing the ability to represent soil erosion and soil transport within the trench cover. Two alternative hypotheses on plant root uptake were incorporated into the model to represent transport of radionuclides by roots that penetrate the buried waste. Enhancements were also made to the scenario for future site intruder activities. Representation of waste package decomposition in the model was confirmed as the best available alternative. Results from sensitivity analyses indicate that additional information is needed to evaluate the alternative hypotheses for plant root uptake of buried wastes. Site-specific evaluations of the contribution from biotic transport to the potential dose to man establish the relevance in the assessment process. The BIOPORT/MAXI1 computer software package is proposed for dose assessments of commercial low-level waste disposal sites.

McKenzie, D.H.; Cadwell, L.L.; Kennedy, W.E. Jr.; Prohammer, L.A.; Simmons, M.A.

1986-11-01T23:59:59.000Z

278

The Mixed Waste Management Facility monthly report, March 1995  

SciTech Connect (OSTI)

This document presents details of the monthly activities of Lawrence Livermore National Laboratory in regards to the Mixed Waste Management Facility. Topics discussed include: quality assurance; regulations; program support; public participation; conceptual design; plant start-up; project management; feed preparation; molten salt, electrochemical, and wet oxidation; process transport and storage; and final waste forms.

Streit, R.D.

1995-04-01T23:59:59.000Z

279

Solid waste management of coal conversion residuals from a commercial-size facility: environmental engineering aspects. Final report  

SciTech Connect (OSTI)

Major residuals generated by the conversion process and its auxiliary operations include: (a) coal preparation wastes; (b) gasifier ash; (c) liquefaction solids-char; (d) tail gas or flue gas desulfurization sludge; (e) boiler flyash and bottom ash; (f) raw water treatment sludge, and; (g) biosludges from process wastewater treatment. Recovered sulfur may also require disposal management. Potential environmental and health impacts from each of the residues are described on the basis of characterization of the waste in the perspective of water quality degradation. Coal gasification and liquefaction systems are described in great detail with respect to their associated residuals. Management options are listed with the conclusion that land disposal of the major residual streams is the only viable choice. On-site versus off-site disposal is analyzed with the selection of on-site operations to reduce political, social and institutional pressures, and to optimize the costs of the system. Mechanisms for prevention of leachate generation are described, and various disposal site designs are outlined. It is concluded that co-disposal feasibility of some waste streams must be established in order to make the most preferred solid waste management system feasible. Capacity requirements for the disposal operation were calculated for a 50,000 bbl/day coal liquefaction plant or 250 million SCF/day gasification operation.

Bern, J.; Neufeld, R. D.; Shapiro, M. A.

1980-11-30T23:59:59.000Z

280

Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers. Final report  

SciTech Connect (OSTI)

One of the most significant factors impacting the performance of waste package container materials under repository relevant conditions is the thermal environment. This environment will be affected by the areal power density of the repository, which is dictated by facility design, and the dominant heat transfer mechanism at the site. The near-field environment will evolve as radioactive decay decreases the thermal output of each waste package. Recent calculations (Buscheck and Nitao, 1994) have addressed the importance of thermal loading conditions on waste package performance at the Yucca Mountain site. If a relatively low repository thermal loading design is employed, the temperature and relative humidity near the waste package may significantly affect the degradation of corrosion allowance barriers due to moist air oxidation and radiolytically enhanced corrosion. The purpose this report is to present a literature review of the potential degradation modes for moderately corrosion resistant nickel copper and nickel based candidate materials that may be applicable as alternate barriers for the ACD systems in the Yucca Mountain environment. This report presents a review of the corrosion of nickel-copper alloys, summaries of experimental evaluations of oxidation and atmospheric corrosion in nickel-copper alloys, views of experimental studies of aqueous corrosion in nickel copper alloys, a brief review of galvanic corrosion effects and a summary of stress corrosion cracking in these alloys.

Vinson, D.W.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

1995-09-22T23:59:59.000Z

Note: This page contains sample records for the topic "final waste form" 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

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

SciTech Connect (OSTI)

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

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

2013-12-01T23:59:59.000Z

282

Final Report - Independent Verification Survey Report for the Waste Loading Area, Former Hazardous Waste Management Facility, Brookhaven National Laboratory, Upton, New York  

SciTech Connect (OSTI)

The objective of the verification survey was to obtain evidence by means of measurements and sampling to confirm that the final radiological conditions were less than the established release criteria. This objective was achieved via multiple verification components including document reviews to determine the accuracy and adequacy of FSS documentation.

P.C. Weaver

2008-08-19T23:59:59.000Z

283

AVS: Experimental Tests of a New Process to Inductively Vitrify HLW Inside the Final Disposal Containers at Very High Waste Loadings  

SciTech Connect (OSTI)

The design and performance capabilities of the Advanced Vitrification System (AVS) are described, together with the results of experimental tests. The AVS is an in-can melting system in which high-level waste (HLW) is vitrified directly inside the final disposal container. The AVS container, or module, consists of an outer stainless steel canister and an alumina-lined, inner graphite crucible, which is thermally insulated from the outer stainless canister. The graphite crucible is inductively heated to very high temperatures (up to 1500 C) by an external low frequency (30 Hertz) alternating current (AC) transformer coil. The actively cooled outer stainless canister remains at near ambient temperature. The HLW/frit mixture is fed into the hot graphite crucible, where it is vitrified. After cooldown, the HLW/frit feed and off-gas pipes are disconnected from the top of the module, which is then sealed and readied for shipment or storage. All radioactively contaminated melter components inside the module are disposed of along with the vitrified waste. The graphite crucible also provides a geologically stable barrier for the vitrified product. The AVS potentially can double HLW loading over that obtained from Joule melters; lower vitrification costs by about half; reduce the number of disposal canisters required by about half; handle diverse waste feeds with high concentrations of problem elements such as chromium and zirconium; and reduce the time needed to vitrify a given inventory of HLW.

Powell, J.; Reich, M.; Jordan, J.; Ventre, L.; Barletta, R.; Manowitz, B.; Steinberg, M.; Grossman, W.; Maise, G.; Salzano, F.; Hess, C.; Ramsey, W. G.; Plodinec, M. J.

2002-02-26T23:59:59.000Z

284

Remaining Sites Verification Package for the 100-B-1 Surface Chemical and Solid Waste Dumping Area, Waste Site Reclassification Form 2006-003  

SciTech Connect (OSTI)

The 100-B-1 waste site was a dumping site that was divided into two areas. One area was used as a laydown area for construction materials, and the other area was used as a chemical dumping area. The 100-B-1 Surface Chemical and Solid Waste Dumping Area site meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

R. A. Carlson

2006-04-24T23:59:59.000Z

285

Final Meeting Summary ...  

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

5 Issue Manager Report-Out on Tank Closure and Waste Management Final Environmental Impact Statement (TC&WM FEIS) Groundwater and Vadose Zone Modeling (JOINT...

286

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

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

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

287

FINAL REPORT U.S. Department of Energy Waste Volume Reduction Using Surface Characterization and Decontamination by  

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

REPORT REPORT U.S. Department of Energy Waste Volume Reduction Using Surface Characterization and Decontamination by Laser Ablation Principal Investigator: Michael J. Pellin Institution: Argonne National Laboratory Collaborators: Michael R, Savina, Claude B. Reed, Zhiyue Xu, Yong Wang Institutions: Argonne National Laboratory Project Number: 60283 Grant Number: W-31-109-ENG-38 Grant Project Officers: C. Miller Project Duration: October 1, 1997 - October 1, 2000 2 Table of Contents Table of Contents............................................................................................................................ 2 Executive Summary ........................................................................................................................ 3 Research Objectives........................................................................................................................

288

Establishment of review groups on US Department of Energy Environmental Restoration and Waste Management Program. Final report  

SciTech Connect (OSTI)

A primary purpose of this grant was the establishment of expert research review groups to help facilitate expanded and improved communications and information among states, public, federal agencies, contractors, and DOE, relative to national environmental and waste management issues/problems. The general objectives of this grant were: Research on the further participation avenues of industry and academia and provide appropriate research documentation concerning the implementation of multi-party agreements; Analysis of the impediments that delay the accomplishment of agreements between states and the federal government for environmental compliance, as well as an assessment of the public need for research because of the above agreements; Analysis of the impact of environmental actions on states, industry, academia, public and other federal agencies; Provide research to help facilitate an interactive system that provides the various involved parties the capability and capacity to strengthen their commitment to national environmental and waste management goals and objectives; and Furthering research of public education in the environmental arena and research of needed national education resources in scientific and technical areas related to environmental restoration and waste management.

Eyman, L.D.

1992-12-01T23:59:59.000Z

289

Tank Waste Disposal Program redefinition  

SciTech Connect (OSTI)

The record of decision (ROD) (DOE 1988) on the Final Environmental Impact Statement, Hanford Defense High-Level, Transuranic and Tank Wastes, Hanford Site, Richland Washington identifies the method for disposal of double-shell tank waste and cesium and strontium capsules at the Hanford Site. The ROD also identifies the need for additional evaluations before a final decision is made on the disposal of single-shell tank waste. This document presents the results of systematic evaluation of the present technical circumstances, alternatives, and regulatory requirements in light of the values of the leaders and constitutents of the program. It recommends a three-phased approach for disposing of tank wastes. This approach allows mature technologies to be applied to the treatment of well-understood waste forms in the near term, while providing time for the development and deployment of successively more advanced pretreatment technologies. The advanced technologies will accelerate disposal by reducing the volume of waste to be vitrified. This document also recommends integration of the double-and single-shell tank waste disposal programs, provides a target schedule for implementation of the selected approach, and describes the essential elements of a program to be baselined in 1992.

Grygiel, M.L.; Augustine, C.A.; Cahill, M.A.; Garfield, J.S.; Johnson, M.E.; Kupfer, M.J.; Meyer, G.A.; Roecker, J.H. [Westinghouse Hanford Co., Richland, WA (United States); Holton, L.K.; Hunter, V.L.; Triplett, M.B. [Pacific Northwest Lab., Richland, WA (United States)

1991-10-01T23:59:59.000Z

290

Microwave solidification development for Rocky Flats waste  

SciTech Connect (OSTI)

The Microwave Engineering Team at the Rocky Flats Plant has developed a production-scale system for the treatment of hazardous, radioactive, and mixed wastes using microwave energy. The system produces a vitreous final form which meets the acceptance criteria for shipment and disposal. The technology also has potential for application on various other waste streams from the public and private sectors. Technology transfer opportunities are being identified and pursued for commercialization of the microwave solidification technology.

Dixon, D.; Erle, R.; Eschen, V. [and others

1994-04-01T23:59:59.000Z

291

Remaining Sites Verification Package for the 100-B-23, 100-B/C Area Surface Debris, Waste Site, Waste Site Reclassification Form 2008-027  

SciTech Connect (OSTI)

The 100-B-23, 100-B/C Surface Debris, waste consisted of multiple locations of surface debris and chemical stains that were identified during an Orphan Site Evaluation of the 100-B/C Area. Evaluation of the collected information for the surface debris features yielded four generic waste groupings: asbestos-containing material, lead debris, oil and oil filters, and treated wood. Focused verification sampling was performed concurrently with remediation. Site remediation was accomplished by selective removal of the suspect hazardous items and potentially impacted soils. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

J. M. Capron

2008-06-16T23:59:59.000Z

292

Tank Waste Committee Page 1  

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

of a PA is to examine the final waste disposition at Hanford, such as waste in the tanks at C-Farm. Vince said the quest is to model waste movement over 10,000 years,...

293

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

294

Environmental assessment of the atlas bio-energy waste wood fluidized bed gasification power plant. Final report  

SciTech Connect (OSTI)

The Atlas Bio-Energy Corporation is proposing to develop and operate a 3 MW power plant in Brooklyn, New York that will produce electricity by gasification of waste wood and combustion of the produced low-Btu gas in a conventional package steam boiler coupled to a steam-electric generator. The objectives of this project were to assist Atlas in addressing the environmental permit requirements for the proposed power plant and to evaluate the environmental and economic impacts of the project compared to more conventional small power plants. The project`s goal was to help promote the commercialization of biomass gasification as an environmentally acceptable and economically attractive alternative to conventional wood combustion. The specific components of this research included: (1) Development of a permitting strategy plan; (2) Characterization of New York City waste wood; (3) Characterization of fluidized bed gasifier/boiler emissions; (4) Performance of an environmental impact analysis; (5) Preparation of an economic evaluation; and (6) Discussion of operational and maintenance concerns. The project is being performed in two phases. Phase I, which is the subject of this report, involves the environmental permitting and environmental/economic assessment of the project. Pending NYSERDA participation, Phase II will include development and implementation of a demonstration program to evaluate the environmental and economic impacts of the full-scale gasification project.

Holzman, M.I.

1995-08-01T23:59:59.000Z

295

Induced Radioactivity and Waste Classification of Reactor Zone Components of the Chernobyl Nuclear Power Plant Unit 1 After Final Shutdown  

SciTech Connect (OSTI)

The dismantlement of the reactor core materials and surrounding structural components is a major technical concern for those planning closure and decontamination and decommissioning of the Chernobyl Nuclear Power Plant (NPP). Specific issues include when and how dismantlement should be accomplished and what the radwaste classification of the dismantled system would be at the time it is disassembled. Whereas radiation levels and residual radiological characteristics of the majority of the plant systems are directly measured using standard radiation survey and radiochemical analysis techniques, actual measurements of reactor zone materials are not practical due to high radiation levels and inaccessibility. For these reasons, neutron transport analysis was used to estimate induced radioactivity and radiation levels in the Chernobyl NPP Unit 1 reactor core materials and structures.Analysis results suggest that the optimum period of safe storage is 90 to 100 yr for the Unit 1 reactor. For all of the reactor components except the fuel channel pipes (or pressure tubes), this will provide sufficient decay time to allow unlimited worker access during dismantlement, minimize the need for expensive remote dismantlement, and allow for the dismantled reactor components to be classified as low- or medium-level radioactive waste. The fuel channel pipes will remain classified as high-activity waste requiring remote dismantlement for hundreds of years due to the high concentration of induced {sup 63}Ni in the Zircaloy pipes.

Bylkin, Boris K. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Davydova, Galina B. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Zverkov, Yuri A. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Krayushkin, Alexander V. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Neretin, Yuri A. [Chernobyl Nuclear Power Plant (Ukraine); Nosovsky, Anatoly V. [Slavutych Division of the International Chernobyl Center (Ukraine); Seyda, Valery A. [Chernobyl Nuclear Power Plant (Ukraine); Short, Steven M. [Pacific Northwest National Laboratory (United States)

2001-10-15T23:59:59.000Z

296

Power plant waste heat utilization in aquaculture. Volume II. Final report, 1 November 1976-1 November 1979  

SciTech Connect (OSTI)

A three-year research study on the constructive use of electric generating station waste heat in cooling effluents for fish production is presented. This volume specifically describes that part of the research conducted by Trenton State College. Water temperatures from the discharge canal of the Mercer Generating Station in New Jersey were blended with those from the Delaware River by pumps installed in strategic locations to achieve desired temperatures. The report further describes how recirculation is controlled during chlorination periods by activating and de-activating certain pumps. As a result of this procedure, plus an oxygen injection system, trout density was greatly increased. Techniques for growing and maintaining shrimp larvae and early juveniles in nursery systems are described. Harvest densities of the shellfish did not compare with those obtained for finfish.

Eble, A.F.

1980-03-01T23:59:59.000Z

297

Waste-heat mariculture of striped bass for population enhancement and food production. Final report on Phase 1  

SciTech Connect (OSTI)

Biological and economic data were obtained to determine whether the culture of striped bass (Morona saxatilis) in power plant thermal effluent could be a cost-effective means of utilizing an otherwise wasted energy resource to provide additional supplies of high-quality seafood. Experiments were conducted to provide some of the data necessary to predict tank-carrying capacity, food-conversion efficiency, and water flow requirements for striped bass cultured at high density in future commercial-scale operations. Computer models were developed for several modes of operation of a theoretical commercial production facility, and return-on-investment calculations were made which indicated that substantial profits are possible. At these sites, no heating or pumping of water would be required, and an annual return-on-operating costs of 103% was estimated.

Van Olst, J.C.; Carlberg, J.M.

1983-04-01T23:59:59.000Z

298

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

SciTech Connect (OSTI)

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

L. M. Dittmer

2006-10-20T23:59:59.000Z

299

Remaining Sites Verification Package for the 128-B-3 Burn Pit Site, Waste Site Reclassification Form 2006-058  

SciTech Connect (OSTI)

The 128-B-3 waste site is a former burn and disposal site for the 100-B/C Area, located adjacent to the Columbia River. The 128-B-3 waste site has been remediated to meet the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results of sampling at upland areas of the site also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2006-11-17T23:59:59.000Z

300

A literature review of mixed waste components: Sensitivities and effects upon solidification/stabilization in cement-based matrices  

SciTech Connect (OSTI)

The US DOE Oak Ridge Field Office has signed a Federal Facility Compliance Agreement (FFCA) regarding Oak Ridge Reservation (ORR) mixed wastes subject to the land disposal restriction (LDR) provisions of the Resource conservation and Recovery Act. The LDR FFCA establishes an aggressive schedule for conducting treatability studies and developing treatment methods for those ORR mixed (radioactive and hazardous) wastes listed in Appendix B to the Agreement. A development, demonstration, testing, and evaluation program has been initiated to provide those efforts necessary to identify treatment methods for all of the wastes that meet Appendix B criteria. The program has assembled project teams to address treatment development needs in a variety of areas, including that of final waste forms (i.e., stabilization/solidification processes). A literature research has been performed, with the objective of determining waste characterization needs to support cement-based waste-form development. The goal was to determine which waste species are problematic in terms of consistent production of an acceptable cement-based waste form and at what concentrations these species become intolerable. The report discusses the following: hydration mechanisms of Portland cement; mechanisms of retardation and acceleration of cement set-factors affecting the durability of waste forms; regulatory limits as they apply to mixed wastes; review of inorganic species that interfere with the development of cement-based waste forms; review of radioactive species that can be immobilized in cement-based waste forms; and review of organic species that may interfere with various waste-form properties.

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

1994-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "final waste form" 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

Development of Technology for Immobilization of Waste Salt from Electrorefining Spent Nuclear Fuel in Zeolite-A for Eventual Disposition in a Ceramic Waste Form  

SciTech Connect (OSTI)

The results of process development for the blending of waste salt from the electrorefining of spent fuel with zeolite-A are presented. This blending is a key step in the ceramic waste process being used for treatment of EBR-II spent fuel and is accomplished using a high-temperature v-blender. A labscale system was used with non-radioactive surrogate salts to determine optimal particle size distributions and time at temperature. An engineering-scale system was then installed in the Hot Fuel Examination Facility hot cell and used to demonstrate blending of actual electrorefiner salt with zeolite. In those tests, it was shown that the results are still favorable with actinide-loaded salt and that batch size of this v-blender could be increased to a level consistent with efficient production operations for EBR-II spent fuel treatment. One technical challenge that remains for this technology is to mitigate the problem of material retention in the v-blender due to formation of caked patches of salt/zeolite on the inner v-blender walls.

Michael F. Simpson; Prateek Sachdev

2008-04-01T23:59:59.000Z

302

Thermal processing system concepts and considerations for RWMC buried waste  

SciTech Connect (OSTI)

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

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

1992-02-01T23:59:59.000Z

303

Development of test acceptance standards for qualification of the glass-bonded zeolite waste form. Interim annual report, October 1995--September 1996  

SciTech Connect (OSTI)

Glass-bonded zeolite is being developed at Argonne National Laboratory in the Electrometallurgical Treatment Program as a potential ceramic waste form for the disposition of radionuclides associated with the US Department of Energy`s (DOE`s) spent nuclear fuel conditioning activities. The utility of standard durability tests [e.g. Materials Characterization Center Test No. 1 (MCC-1), Product Consistency Test (PCT), and Vapor Hydration Test (VHT)] are being evaluated as an initial step in developing test methods that can be used in the process of qualifying this material for acceptance into the Civilian Radioactive Waste Management System. A broad range of potential repository conditions are being evaluated to determine the bounding parameters appropriate for the corrosion testing of the ceramic waste form, and its behavior under accelerated testing conditions. In this report we provide specific characterization information and discuss how the durability test results are affected by changes in pH, leachant composition, and sample surface area to leachant volume ratios. We investigate the release mechanisms and other physical and chemical parameters that are important for establishing acceptance parameters, including the development of appropriate test methodologies required to measure product consistency.

Simpson, L.J.; Wronkiewicz, D.J.; Fortner, J.A.

1997-09-01T23:59:59.000Z

304

System Engineering Design [Nuclear Waste Management using  

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

System Engineering System Engineering Design Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Waste Management using Electrometallurgical Technology System Engineering Design Bookmark and Share Two major pieces of electrometallurgical process equipment are the Electrorefiner and the Cathode Processor. NE personnel have been involved in the conceptual design, final design, procurement, manufacture,

305

Remaining Sites Verification Package for the 128-F-2, 100-F Burning Pit Waste Site, Waste Site Reclassification Form 2008-031  

SciTech Connect (OSTI)

The 128-F-2 waste site consisted of multiple burn and debris filled pits located directly east of the 107-F Retention Basin and approximately 30.5 m east of the northeast corner of the 100-F Area perimeter road that runs along the riverbank. The burn pits were used for incinerating nonradioactive, combustible materials from 1945 to 1965. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

J. M. Capron

2008-12-01T23:59:59.000Z

306

Final Environmental Impact Statement for Treating Transuranic (TRU)/Alpha Low-level Waste at the Oak Ridge National Laboratory Oak Ridge, Tennessee  

SciTech Connect (OSTI)

The DOE proposes to construct, operate, and decontaminate/decommission a TRU Waste Treatment Facility in Oak Ridge, Tennessee. The four waste types that would be treated at the proposed facility would be remote-handled TRU mixed waste sludge, liquid low-level waste associated with the sludge, contact-handled TRU/alpha low-level waste solids, and remote-handled TRU/alpha low-level waste solids. The mixed waste sludge and some of the solid waste contain metals regulated under the Resource Conservation and Recovery Act and may be classified as mixed waste. This document analyzes the potential environmental impacts associated with five alternatives--No Action, the Low-Temperature Drying Alternative (Preferred Alternative), the Vitrification Alternative, the Cementation Alternative, and the Treatment and Waste Storage at Oak Ridge National Laboratory (ORNL) Alternative.

N /A

2000-06-30T23:59:59.000Z

307

Nuclear Waste  

Science Journals Connector (OSTI)

Nuclear waste is radioactive material no longer considered valuable...238U, 235U, and 226Ra (where the latter decays to 222Rn gas by emitting an alpha particle) or formed through fission of fissile radioisotopes ...

Rob P. Rechard

2014-01-01T23:59:59.000Z

308

Heavy-Duty Waste Hauler with Chemically Correct Natural Gas Engine Diluted with EGR and Using a Three-Way Catalyst: Final Report, 24 February 2004 -- 23 February 2006  

Alternative Fuels and Advanced Vehicles Data Center [Office of Energy Efficiency and Renewable Energy (EERE)]

Heavy-Duty Waste Hauler with Heavy-Duty Waste Hauler with Chemically Correct Natural Gas Engine Diluted with EGR and Using a Three-Way Catalyst Final Report February 24, 2004 - February 23, 2006 T. Reppert Mack Trucks, Inc. Allentown, Pennsylvania J. Chiu Southwest Research Institute San Antonio, Texas Subcontract Report NREL/SR-540-38222 September 2005 Heavy-Duty Waste Hauler with Chemically Correct Natural Gas Engine Diluted with EGR and Using a Three-Way Catalyst Final Report February 24, 2004 - February 23, 2006 T. Reppert Mack Trucks, Inc. Allentown, Pennsylvania J. Chiu Southwest Research Institute San Antonio, Texas NREL Technical Monitor: R. Parish Prepared under Subcontract No. ZCI-4-32049-01 Subcontract Report NREL/SR-540-38222 September 2005 National Renewable Energy Laboratory

309

Commercialization of waste gob gas and methane produced in conjunction with coal mining operations. Final report, August 1992--December 1993  

SciTech Connect (OSTI)

The primary objectives of the project were to identify and evaluate existing processes for (1) using gas as a feedstock for production of marketable, value-added commodities, and (2) enriching contaminated gas to pipeline quality. The following gas conversion technologies were evaluated: (1) transformation to liquid fuels, (2) manufacture of methanol, (3) synthesis of mixed alcohols, and (4) conversion to ammonia and urea. All of these involved synthesis gas production prior to conversion to the desired end products. Most of the conversion technologies evaluated were found to be mature processes operating at a large scale. A drawback in all of the processes was the need to have a relatively pure feedstock, thereby requiring gas clean-up prior to conversion. Despite this requirement, the conversion technologies were preliminarily found to be marginally economic. However, the prohibitively high investment for a combined gas clean-up/conversion facility required that REI refocus the project to investigation of gas enrichment alternatives. Enrichment of a gas stream with only one contaminant is a relatively straightforward process (depending on the contaminant) using available technology. However, gob gas has a unique nature, being typically composed of from constituents. These components are: methane, nitrogen, oxygen, carbon dioxide and water vapor. Each of the four contaminants may be separated from the methane using existing technologies that have varying degrees of complexity and compatibility. However, the operating and cost effectiveness of the combined system is dependent on careful integration of the clean-up processes. REI is pursuing Phase 2 of this project for demonstration of a waste gas enrichment facility using the approach described above. This is expected to result in the validation of the commercial and technical viability of the facility, and the refinement of design parameters.

Not Available

1993-12-01T23:59:59.000Z

310

Screening study for waste biomass to ethanol production facility using the Amoco process in New York State. Final report  

SciTech Connect (OSTI)

This report evaluates the economic feasibility of locating biomass-to-ethanol waste conversion facilities in New York State. Part 1 of the study evaluates 74 potential sites in New York City and identifies two preferred sites on Staten, the Proctor Gamble and the Arthur Kill sites, for further consideration. Part 2 evaluates upstate New York and determines that four regions surrounding the urban centers of Albany, Buffalo, Rochester, and Syracuse provide suitable areas from which to select specific sites for further consideration. A separate Appendix provides supplemental material supporting the evaluations. A conceptual design and economic viability evaluation were developed for a minimum-size facility capable of processing 500 tons per day (tpd) of biomass consisting of wood or paper, or a combination of the two for upstate regions. The facility would use Amoco`s biomass conversion technology and produce 49,000 gallons per day of ethanol and approximately 300 tpd of lignin solid by-product. For New York City, a 1,000-tpd processing facility was also evaluated to examine effects of economies of scale. The reports evaluate the feasibility of building a biomass conversion facility in terms of city and state economic, environmental, and community factors. Given the data obtained to date, including changing costs for feedstock and ethanol, the project is marginally attractive. A facility should be as large as possible and located in a New York State Economic Development Zone to take advantage of economic incentives. The facility should have on-site oxidation capabilities, which will make it more financially viable given the high cost of energy. 26 figs., 121 tabs.

NONE

1995-08-01T23:59:59.000Z

311

Final Report: Modifications and Optimization of the Organic Rankine Cycle to Improve the Recovery of Waste Heat  

SciTech Connect (OSTI)

This research and development (R&D) project exemplifies a shared public private commitment to advance the development of energy efficient industrial technologies that will reduce the U.S. dependence upon foreign oil, provide energy savings and reduce greenhouse gas emissions. The purpose of this project was to develop and demonstrate a Direct Evaporator for the Organic Rankine Cycle (ORC) for the conversion of waste heat from gas turbine exhaust to electricity. In conventional ORCs, the heat from the exhaust stream is transferred indirectly to a hydrocarbon based working fluid by means of an intermediate thermal oil loop. The Direct Evaporator accomplishes preheating, evaporation and superheating of the working fluid by a heat exchanger placed within the exhaust gas stream. Direct Evaporation is simpler and up to 15% less expensive than conventional ORCs, since the secondary oil loop and associated equipment can be eliminated. However, in the past, Direct Evaporation has been avoided due to technical challenges imposed by decomposition and flammability of the working fluid. The purpose of this project was to retire key risks and overcome the technical barriers to implementing an ORC with Direct Evaporation. R&D was conducted through a partnership between the Idaho National Laboratory (INL) and General Electric (GE) Global Research Center (GRC). The project consisted of four research tasks: (1) Detailed Design & Modeling of the ORC Direct Evaporator, (2) Design and Construction of Partial Prototype Direct Evaporator Test Facility, (3) Working Fluid Decomposition Chemical Analyses, and (4) Prototype Evaluation. Issues pertinent to the selection of an ORC working fluid, along with thermodynamic and design considerations of the direct evaporator, were identified. The FMEA (Failure modes and effects analysis) and HAZOP (Hazards and operability analysis) safety studies performed to mitigate risks are described, followed by a discussion of the flammability analysis of the direct evaporator. A testbed was constructed and the prototype demonstrated at the GE GRC Niskayuna facility.

Donna Post Guillen; Jalal Zia

2013-09-01T23:59:59.000Z

312

Remaining Sites Verification Package for the 1607-F4 Sanitary Sewer System, Waste Site Reclassification Form 2004-131  

SciTech Connect (OSTI)

The 1607-F4 waste site is the former location of the sanitary sewer system that serviced the former 115-F Gas Recirculation Building. The system included a septic tank, drain field, and associated pipeline that were in use from 1944 to 1965. The 1607-F4 waste site received unknown amounts of sanitary sewage from the 115-F Gas Recirculation Building and may have potentially contained hazardous and radioactive contamination. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2007-12-03T23:59:59.000Z

313

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

SciTech Connect (OSTI)

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

NONE

1995-02-01T23:59:59.000Z

314

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

315

EIS-0026-S: Final Supplemental Environmental Impact Statement  

Broader source: Energy.gov [DOE]

Final Supplemental Environmental Impact Statement, Waste Isolation Pilot Plant (WIPP), Carlsbad, New Mexico

316

Flue gas desulfurization sludge: establishment of vegetation on ponded and soil-applied waste. Final report January 1977-September 1981  

SciTech Connect (OSTI)

The report gives results of research to identify and evaluate forms of vegetation and methods of their establishment for reclaiming retired flue gas desulfurization sludge ponds. Also studied were the soil liming value of limestone scrubber sludge (LSS) and plant uptake and percolation losses of some chemical nutrients in the sludge. Several vegetation schemes were evaluated between 1977 and 1982 for covering and stabilizing LSS at Colbert Steam Plant, Cherokee, AL, and Shawnee Steam Plant, Paducah, KY. Eleven tree and 10 grass or legume species were tested for adaptability and survival when planted directly in LSS or in LSS amended with soil, municipal sewage sludge, or standard potting mix. Other studies indicated that LSS apparently has sufficient unreacted limestone to be a satisfactory soil liming agent.

Giordano, P.M.; Mays, D.A.; Soileau, J.M.

1984-01-01T23:59:59.000Z

317

Final Report - High Level Waste Vitrification System Improvements, VSL-07R1010-1, Rev 0, dated 04/16/07  

SciTech Connect (OSTI)

This report describes work conducted to support the development and testing of new glass formulations that extend beyond those that have been previously investigated for the Hanford Waste Treatment and Immobilization Plant (WTP). The principal objective was to investigate maximization of the incorporation of several waste components that are expected to limit waste loading and, consequently, high level waste (HLW) processing rates and canister count. The work was performed with four waste compositions specified by the Office of River Protection (ORP); these wastes contain high concentrations of bismuth, chromium, aluminum, and aluminum plus sodium. The tests were designed to identify glass formulations that maximize waste loading while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass formulations, increased glass processing temperature, increased crystallinity, and feed solids content on waste processing rate and product quality.

Kruger, Albert A.; Gan, H.; Pegg, I. L.; Gong, W.; Champman, C. C.; Joseph, I.; Matlack, K. S.

2013-11-13T23:59:59.000Z

318

Remaining Sites Verification Package for the 1607-F3 Sanitary Sewer System, Waste Site Reclassification Form 2006-047  

SciTech Connect (OSTI)

The 1607-F3 waste site is the former location of the sanitary sewer system that supported the 182-F Pump Station, the 183-F Water Treatment Plant, and the 151-F Substation. The sanitary sewer system included a septic tank, drain field, and associated pipeline, all in use between 1944 and 1965. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2007-04-26T23:59:59.000Z

319

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

SciTech Connect (OSTI)

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

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

2011-06-08T23:59:59.000Z

320

EIS-0026: Final Environmental Impact Statement | Department of...  

Energy Savers [EERE]

26: Final Environmental Impact Statement EIS-0026: Final Environmental Impact Statement Waste Isolation Pilot Plant (WIPP), Carlsbad, New Mexico EIS-0026: Final Environmental...

Note: This page contains sample records for the topic "final waste form" 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

EIS-0290: Final Environmental Impact Statement | Department of...  

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

Final Environmental Impact Statement EIS-0290: Final Environmental Impact Statement Advanced Mixed Waste Treatment Project (AMWTP) The AMWTP Final EIS assesses the potential...

322

Single Pass Flow-Through (SPFT) Test Results of Fluidized Bed Steam Reforming (FBSR) Waste Forms used for LAW Immobilization  

SciTech Connect (OSTI)

Several supplemental technologies for treating and immobilizing Hanford low activity waste (LAW) are being evaluated. One such immobilization technology being considered is the Fluidized Bed Steam Reforming (FBSR) granular product. The FBSR granular product is composed of insoluble sodium aluminosilicate (NAS) feldspathoid minerals. Production of the FBSR mineral product has been demonstrated both at the industrial and laboratory scale. Single-Pass Flow-Through (SPFT) tests at various flow rates have been conducted with the granular products fabricated using these two methods. Results show that the materials exhibit a relatively low forward dissolution rate on the order of 10-3 g/(m2d) with the material made in the laboratory giving slightly higher values.

Neeway, James J.; Qafoku, Nikolla; Williams, Benjamin D.; Valenta, Michelle M.; Cordova, Elsa A.; Strandquist, Sara C.; Dage, DeNomy C.; Brown, Christopher F.

2012-03-20T23:59:59.000Z

323

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

SciTech Connect (OSTI)

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

R. A. Carlson

2006-05-04T23:59:59.000Z

324

Remaining Sites Verification Package for the 1607-B1 Septic System, Waste Site Reclassification Form 2007-015  

SciTech Connect (OSTI)

The 1607-B1 Septic System includes a septic tank, drain field, and associated connecting pipelines and influent sanitary sewer lines. This septic system serviced the former 1701-B Badgehouse, 1720-B Patrol Building/Change Room, and the 1709-B Fire Headquarters. The 1607-B1 waste site received unknown amounts of nonhazardous, nonradioactive sanitary sewage from these facilities during its operational history from 1944 to approximately 1970. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2007-08-30T23:59:59.000Z

325

Final Meeting Summary Page 1  

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

October 8, 2014 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD TANK WASTE COMMITTEE October 8, 2014 Richland, WA Topics in this Meeting Summary Opening ......

326

Nuclear waste solids  

Science Journals Connector (OSTI)

Glass and polycrystalline materials for high-level radioactive waste immobilization are discussed. Borosilicate glass has been selected as the waste form for defence high-level radwaste in the US. Since releas...

L. L. Hench; D. E. Clark; A. B. Harker

1986-05-01T23:59:59.000Z

327

Testing of low-temperature stabilization alternatives for salt containing mixed wastes -- Approach and results to date  

SciTech Connect (OSTI)

Through its annual process of identifying technology deficiencies associated with waste treatment, the Department of Energy`s (DOE) Mixed Waste Focus Area (MWFA) determined that the former DOE weapons complex lacks efficient mixed waste stabilization technologies for salt containing wastes. These wastes were generated as sludge and solid effluents from various primary nuclear processes involving acids and metal finishing; and well over 10,000 cubic meters exist at 6 sites. In addition, future volumes of these problematic wastes will be produced as other mixed waste treatment methods such as incineration and melting are deployed. The current method used to stabilize salt waste for compliant disposal is grouting with Portland cement. This method is inefficient since the highly soluble and reactive chloride, nitrate, and sulfate salts interfere with the hydration and setting processes associated with grouting. The inefficiency results from having to use low waste loadings to ensure a durable and leach resistant final waste form. The following five alternatives were selected for MWFA development funding in FY97 and FY98: phosphate bonded ceramics; sol-gel process; polysiloxane; polyester resin; and enhanced concrete. Comparable evaluations were planned for the stabilization development efforts. Under these evaluations each technology stabilized the same type of salt waste surrogates. Final waste form performance data such as compressive strength, waste loading, and leachability could then be equally compared. Selected preliminary test results are provided in this paper.

Maio, V.; Loomis, G. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Spence, R.D. [Oak Ridge National Lab., TN (United States); Smith, G. [Pacific Northwest National Lab., Richland, WA (United States); Biyani, R.K. [SGN Eurisys Services Corp., Richland, WA (United States); Wagh, A. [Argonne National Lab., IL (United States)

1998-05-01T23:59:59.000Z

328

Chemically bonded phosphate ceramics for low-level mixed waste stabilization  

SciTech Connect (OSTI)

Novel chemically bonded phosphate ceramics (CBPCs) are being developed and fabricated for low-temperature stabilization and solidification of mixed waste streams which are amenable to conventional high-temperature stabilization processes due to presence of volatiles such as heavy metal chloride and fluorides and/or pyrophorics in the wastes. Phosphates of Mg, Mg-Na and Zr are being developed as candidate matrix materials. In this paper, we present the fabrication procedures of phosphate waste forms using surrogates compositions of three typical mixed wastes streams -- ash, cement sludges, and salts. The performance of the final waste forms such as compression strength, leachability of the contaminants, durability in aqueous environment were conducted. In addition, parameteric studies have been conducted to establish the optimal waste loading in a particular binder system. Based on the results, we present potential applications in the treatment of various mixed waste streams.

Singh, D.; Wagh, A.S.; Cunnane, J.C. [Argonne National Lab., IL (United States); Mayberry, J.L. [Science Applications International Corp., Idaho Falls, ID (United States)

1994-12-31T23:59:59.000Z

329

EIS-0287: Idaho High-Level Waste & Facilities Disposition | Department of  

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

7: Idaho High-Level Waste & Facilities Disposition 7: Idaho High-Level Waste & Facilities Disposition EIS-0287: Idaho High-Level Waste & Facilities Disposition SUMMARY This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liquid waste at the Idaho National Engineering and Environmental Laboratory (INEEL) in liquid and solid forms. This EIS also analyzes alternatives for the final disposition of HLW management facilities at the INEEL after their missions are completed. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD January 12, 2010 EIS-0287: Amended Record of Decision Idaho High-Level Waste and Facilities Disposition January 4, 2010

330

Effects of heat treatment and formulation on the phase composition and chemical durability of the EBR-ll ceramic waste form.  

SciTech Connect (OSTI)

High-level radioactive waste salts generated during the electrometallurgical treatment of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor-II will be immobilized in a ceramic waste form (CWF). Tests are being conducted to evaluate the suitability of the CWF for disposal in the planned federal high-level radioactive waste repository at Yucca Mountain. In this report, the results of laboratory tests and analyses conducted to address product consistency and thermal stability issues called out in waste acceptance requirements are presented. The tests measure the impacts of (1) variations in the amounts of salt and binder glass used to make the CWF and (2) heat treatments on the phase composition and chemical durability of the waste form. A series of CWF materials was made to span the ranges of salt and glass contents that could be used during processing: between 5.0 and 15 mass% salt loaded into the zeolite (the nominal salt loading is 10.7%, and the process control range is 10.6 to 11.2 mass%), and between 20 and 30 mass% binder glass mixed with the salt-loaded zeolite (the nominal glass content is 25% and the process control range is 20 to 30 mass%). In another series of tests, samples of two CWF products made with the nominal salt and glass contents were reheated to measure the impact on the phase composition and durability: long-term heat treatments were conducted at 400 and 500 C for durations of 1 week, 4 weeks, 3 months, 6 months, and 1 year; short-term heat treatments were conducted at 600, 700, 800, and 850 C for durations of 4, 28, 52, and 100 hours. All of the CWF products that were made with different amounts of salt, zeolite, and glass and all of the heat-treated CWF samples were analyzed with powder X-ray diffraction to measure changes in phase compositions and subjected to 7-day product consistency tests to measure changes in the chemical durability. The salt loading had the greatest impact on phase composition and durability. A relatively large amount of nepheline, Na{sub 4}(AlSiO{sub 4}){sub 4}, was formed in the material made with 5.0 mass% salt loading, which was also the least durable of the materials that were tested. Nepheline was not detected in materials made with salt-loaded zeolites containing 15 or 20 mass% salt. Conversely, halite was not detected with XRD in materials made with 5.0 or 7.5 mass% salt loading, but similar amounts of halite were measured in the other CWF materials. The sodalite contents of all materials were similar. The halite content in the CWF source material used in the short-term heat-treatment study, which had the nominal salt and binder glass loadings, was determined to be about 1.3 mass% by standard addition analysis. Heat treatment had only a small effect on the phase composition: the amount of halite increased to as much as 3.7 mass%, and trace amounts of nepheline were detected in samples treated at 800 and 850 C. The CWF samples treated at high temperatures had lower amounts of halite detected in the rapid water-soluble test. The releases of B, Na, and Si in the product consistency tests (PCTs) were not sensitive to the heat-treatment conditions. The PCT responses of all salt-loaded and heat-treated CWF materials were well below that of the Environmental Assessment (EA) glass.

Ebert, W. E.; Dietz, N. L.; Janney, D. E.

2006-01-31T23:59:59.000Z

331

Characteristics of potential repository wastes. Volume 3, Appendix 3A, ORIGEN2 decay tables for immobilized high-level waste; Appendix 3B, Interim high-level waste forms  

SciTech Connect (OSTI)

This appendix presents the results of decay calculations using the ORIGEN2 code to determine the radiological properties of canisters of immobilized high-level waste as a function of decay time for decay times up to one million years. These calculations were made for the four HLW sites (West Valley Demonstration Project, Savannah River Site, Hanford Site, and Idaho National Engineering Laboratory) using the composition data discussed in the HLW section of this report. Calculated ({alpha},n) neutron production rates are also shown.

Not Available

1992-07-01T23:59:59.000Z

332

Draft forecast of the final report for the comparison to 40 CFR Part 191, Subpart B, for the Waste Isolation Pilot Plant  

SciTech Connect (OSTI)

The United States Department of Energy is planning to dispose of transuranic wastes, which have been generated by defense programs, at the Waste Isolation Pilot Plant. The WIPP Project will assess compliance with the requirements of the United States Environmental Protection Agency. This report forecasts the planned 1992 document, Comparison to 40 CFR, Part 191, Subpart B, for the Waste Isolation Pilot Plant (WIPP). 130 refs., 36 figs., 11 tabs.

Bertram-Howery, S.G.; Marietta, M.G.; Anderson, D.R.; Gomez, L.S.; Rechard, R.P. (Sandia National Labs., Albuquerque, NM (USA)); Brinster, K.F.; Guzowski, R.V. (Science Applications International Corp., Albuquerque, NM (USA))

1989-12-01T23:59:59.000Z

333

Radioactive Demonstrations Of Fluidized Bed Steam Reforming (FBSR) With Hanford Low Activity Wastes  

SciTech Connect (OSTI)

Several supplemental technologies for treating and immobilizing Hanford low activity waste (LAW) are being evaluated. One immobilization technology being considered is Fluidized Bed Steam Reforming (FBSR) which offers a low temperature (700-750?C) continuous method by which wastes high in organics, nitrates, sulfates/sulfides, or other aqueous components may be processed into a crystalline ceramic (mineral) waste form. The granular waste form produced by co-processing the waste with kaolin clay has been shown to be as durable as LAW glass. The FBSR granular product will be monolithed into a final waste form. The granular component is composed of insoluble sodium aluminosilicate (NAS) feldspathoid minerals such as sodalite. Production of the FBSR mineral product has been demonstrated both at the industrial, engineering, pilot, and laboratory scales on simulants. Radioactive testing at SRNL commenced in late 2010 to demonstrate the technology on radioactive LAW streams which is the focus of this study.

Jantzen, C. M.; Crawford, C. L.; Burket, P. R.; Bannochie, C. J.; Daniel, W. G.; Nash, C. A.; Cozzi, A. D.; Herman, C. C.

2012-10-22T23:59:59.000Z

334

Hawaii Permit Application for Solid Waste Management Facility...  

Open Energy Info (EERE)

to receive a permit for a solid waste management facility. Form Type CertificateForm of Completion Form Topic Permit Application for Solid Waste Management Facility Organization...

335

Control of water infiltration into near surface low-level waste disposal units. Final report on field experiments at a humid region site, Beltsville, Maryland  

SciTech Connect (OSTI)

This study`s objective was to assess means for controlling water infiltration through waste disposal unit covers in humid regions. Experimental work was carried out in large-scale lysimeters 21.34 m x 13.72 m x 3.05 m (70 ft x 45 ft x 10 ft) at Beltsville, Maryland. Results of the assessment are applicable to disposal of low-level radioactive waste (LLW), uranium mill tailings, hazardous waste, and sanitary landfills. Three kinds of waste disposal unit covers or barriers to water infiltration were investigated: (1) resistive layer barrier, (2) conductive layer barrier, and (3) bioengineering management.

Schulz, R.K.; Ridky, R.W.; O`Donnell, E.

1997-09-01T23:59:59.000Z

336

Melter development needs assessment for RWMC buried wastes  

SciTech Connect (OSTI)

This report presents a survey and initial assessment of the existing state-of-the-art melter technology necessary to thermally treat (stabilize) buried TRU waste, by producing a highly leach resistant glass/ceramic waste form suitable for final disposal. Buried mixed transuranic (TRU) waste at the Idaho National Engineering Laboratory (INEL) represents an environmental hazard requiring remediation. The Environmental Protection Agency (EPA) placed the INEL on the National Priorities List in 1989. Remediation of the buried TRU-contaminated waste via the CERCLA decision process is required to remove INEL from the National Priorities List. A Waste Technology Development (WTD) Preliminary Systems Design and Thermal Technologies Screening Study identified joule-heated and plasma-heated melters as the most probable thermal systems technologies capable of melting the INEL soil and waste to produce the desired final waste form (Iron-Enriched Basalt (IEB) glass/ceramic). The work reported herein then surveys the state of existing melter technology and assesses it within the context of processing INEL buried TRU wastes and contaminated soils. Necessary technology development work is recommended.

Donaldson, A.D.; Carpenedo, R.J.; Anderson, G.L.

1992-02-01T23:59:59.000Z

337

Melter development needs assessment for RWMC buried wastes  

SciTech Connect (OSTI)

This report presents a survey and initial assessment of the existing state-of-the-art melter technology necessary to thermally treat (stabilize) buried TRU waste, by producing a highly leach resistant glass/ceramic waste form suitable for final disposal. Buried mixed transuranic (TRU) waste at the Idaho National Engineering Laboratory (INEL) represents an environmental hazard requiring remediation. The Environmental Protection Agency (EPA) placed the INEL on the National Priorities List in 1989. Remediation of the buried TRU-contaminated waste via the CERCLA decision process is required to remove INEL from the National Priorities List. A Waste Technology Development (WTD) Preliminary Systems Design and Thermal Technologies Screening Study identified joule-heated and plasma-heated melters as the most probable thermal systems technologies capable of melting the INEL soil and waste to produce the desired final waste form [Iron-Enriched Basalt (IEB) glass/ceramic]. The work reported herein then surveys the state of existing melter technology and assesses it within the context of processing INEL buried TRU wastes and contaminated soils. Necessary technology development work is recommended.

Donaldson, A.D.; Carpenedo, R.J.; Anderson, G.L.

1992-02-01T23:59:59.000Z

338

Development of an Integrated Raman and Turbidity Fiber Optic Sensor for the In-Situ Analysis of High Level Nuclear Waste - 13532  

SciTech Connect (OSTI)

Stored nuclear waste must be retrieved from storage, treated, separated into low- and high-level waste streams, and finally put into a disposal form that effectively encapsulates the waste and isolates it from the environment for a long period of time. Before waste retrieval can be done, waste composition needs to be characterized so that proper safety precautions can be implemented during the retrieval process. In addition, there is a need for active monitoring of the dynamic chemistry of the waste during storage since the waste composition can become highly corrosive. This work describes the development of a novel, integrated fiber optic Raman and light scattering probe for in situ use in nuclear waste solutions. The dual Raman and turbidity sensor provides simultaneous chemical identification of nuclear waste as well as information concerning the suspended particles in the waste using a common laser excitation source. (authors)

Gasbarro, Christina; Bello, Job [EIC Laboratories, Inc., 111 Downey St., Norwood, MA, 02062 (United States)] [EIC Laboratories, Inc., 111 Downey St., Norwood, MA, 02062 (United States); Bryan, Samuel; Lines, Amanda; Levitskaia, Tatiana [Pacific Northwest National Laboratory, PO Box 999, Richland, WA, 99352 (United States)] [Pacific Northwest National Laboratory, PO Box 999, Richland, WA, 99352 (United States)

2013-07-01T23:59:59.000Z

339

Radioactive Waste Radioactive Waste  

E-Print Network [OSTI]

#12;Radioactive Waste at UF Bldg 831 392-8400 #12;Radioactive Waste · Program is designed to;Radioactive Waste · Program requires · Generator support · Proper segregation · Packaging · labeling #12;Radioactive Waste · What is radioactive waste? · Anything that · Contains · or is contaminated

Slatton, Clint

340

Disposal of Greater-than-Class C Low-Level Radioactive Waste  

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

Disposal of Low-Level Radioactive Waste Disposal of Low-Level Radioactive Waste EVS prepared a draft environmental impact statement (EIS) for disposal of greater-than-Class C low-level radioactive waste (GTCC LLRW). The EVS Division prepared a draft environmental impact statement (EIS) for disposal of greater-than-Class C low-level radioactive waste (GTCC LLRW) for the DOE Office of Environmental Management. DOE is now finalizing this EIS and is including a preferred alternative. DOE intends that the final EIS will provide information to support the selection of disposal method(s) and site(s) for GTCC LLRW and GTCC-like waste. In general, GTCC LLRW is not acceptable for near-surface disposal. Typically, the waste form and disposal methods must be different from and more stringent than those specified for Class C LLRW. For GTCC LLRW, the

Note: This page contains sample records for the topic "final waste form" 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

Glassy slags for minimum additive waste stabilization. Interim progress report, May 1993--February 1994  

SciTech Connect (OSTI)

Glassy slag waste forms are being developed to complement glass waste forms in implementing Minimum Additive Waste Stabilization (MAWS) for supporting DOE`s environmental restoration efforts. The glassy slag waste form is composed of various crystalline and metal oxide phases embedded in a silicate glass phase. The MAWS approach was adopted by blending multiple waste streams to achieve up to 100% waste loadings. The crystalline phases, such as spinels, are very durable and contain hazardous and radioactive elements in their lattice structures. These crystalline phases may account for up to 80% of the total volume of slags having over 80% metal loading. The structural bond strength model was used to quantify the correlation between glassy slag composition and chemical durability so that optimized slag compositions were obtained with limited crucible melting and testing. Slag compositions developed through crucible melts were also successfully generated in a pilot-scale Retech plasma centrifugal furnace at Ukiah, California. Utilization of glassy slag waste forms allows the MAWS approach to be applied to a much wider range of waste streams than glass waste forms. The initial work at ANL has indicated that glassy slags are good final waste forms because of (1) their high chemical durability; (2) their ability to incorporate large amounts of metal oxides; (3) their ability to incorporate waste streams having low contents of flux components; (4) their less stringent requirements on processing parameters, compared to glass waste forms; and (5) their low requirements for purchased additives, which means greater waste volume reduction and treatment cost savings.

Feng, X.; Wronkiewicz, D.J.; Bates, J.K.; Brown, N.R.; Buck, E.C.; Dietz, N.L.; Gong, M.; Emery, J.W. [Argonne National Lab., IL (United States). Chemical Technology Div.] [Argonne National Lab., IL (United States). Chemical Technology Div.

1994-05-01T23:59:59.000Z

342

Central Waste Complex (CWC) Waste Analysis Plan  

SciTech Connect (OSTI)

The purpose of this waste analysis plan (WAP) is to document the waste acceptance process, sampling methodologies, analytical techniques, and overall processes that are undertaken for waste accepted for storage at the Central Waste Complex (CWC), which is located in the 200 West Area of the Hanford Facility, Richland, Washington. Because dangerous waste does not include the source special nuclear and by-product material components of mixed waste, radionuclides are not within the scope of this document. The information on radionuclides is provided only for general knowledge. This document has been revised to meet the interim status waste analysis plan requirements of Washington Administrative Code (WAC) 173 303-300(5). When the final status permit is issued, permit conditions will be incorporated and this document will be revised accordingly.

ELLEFSON, M.D.

2000-01-06T23:59:59.000Z

343

DOE Issues Final Request for Proposal for Oak Ridge Transuranic...  

Office of Environmental Management (EM)

Oak Ridge Transuranic Waste Processing Center Services DOE Issues Final Request for Proposal for Oak Ridge Transuranic Waste Processing Center Services October 14, 2014 - 2:16pm...

344

Safety Analysis, Hazard and Risk Evaluations [Nuclear Waste Management  

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

Safety Analysis, Hazard Safety Analysis, Hazard and Risk Evaluations Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Waste Management using Electrometallurgical Technology Safety Analysis, Hazard and Risk Evaluations Bookmark and Share NE Division personnel had a key role in the creation of the FCF Final Safety Analysis Report (FSAR), FCF Technical Safety Requirements (TSR)

345

Hydration Aging of Nuclear Waste Glass  

Science Journals Connector (OSTI)

...of Nuclear Waste Glass 10...STEINDLER Chemical Engineering...60439 The aging of simulated nuclear waste glass by...nuclear waste forms can meet...simulated aging reac-tions...whether a waste formn can...pro-jected Nuclear Regulatory...STEINDLEt Chemical Engineering...Basisfor Waste Form Integrity...

J. K. BATES; L. J. JARDINE; M. J. STEINDLER

1982-10-01T23:59:59.000Z

346

Waste disposal package  

DOE Patents [OSTI]

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

Smith, M.J.

1985-06-19T23:59:59.000Z

347

Long-range master plan for defense transuranic waste management  

SciTech Connect (OSTI)

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

Not Available

1988-12-01T23:59:59.000Z

348

Tank Waste Committee Page 1  

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

7, 2014 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD TANK WASTE COMMITTEE May 7, 2014 Richland, WA Topics in this Meeting Summary Opening ......

349

Tank Waste Committee Page 1  

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

June 9, 2011 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD TANK WASTE COMMITTEE MEETING June 9, 2011 Richland, WA Topics in this Meeting Summary Welcome and Introductions...

350

Treatability studies of actual listed waste sludges from the Oak Ridge Reservation (ORR)  

SciTech Connect (OSTI)

Oak Ridge National Laboratory (ORNL) and Savannah River Technology Center (SRTC) are investigating vitrification for various low-level and mixed wastes on the Oak Ridge Reservation (ORR). Treatability studies have included surrogate waste formulations at the laboratory-, pilot-, and field-scales and actual waste testing at the laboratory- and pilot-scales. The initial waste to be processing through SRTC`s Transportable Vitrification System (TVS) is the K-1407-B and K-1407-C (B/C) Pond sludge waste which is a RCRA F-listed waste. The B/C ponds at the ORR K-25 site were used as holding and settling ponds for various waste water treatment streams. Laboratory-, pilot-, and field- scale ``proof-of-principle`` demonstrations are providing needed operating parameters for the planned field-scale demonstration with actual B/C Pond sludge waste at ORR. This report discusses the applied systems approach to optimize glass compositions for this particular waste stream through laboratory-, pilot-, and field-scale studies with surrogate and actual B/C waste. These glass compositions will maximize glass durability and waste loading while optimizing melt properties which affect melter operation, such as melt viscosity and melter refractory corrosion. Maximum waste loadings minimize storage volume of the final waste form translating into considerable cost savings.

Jantzen, C.M.; Peeler, D.K. [Westinghouse Savannah River Co., Aiken, SC (United States); Gilliam, T.M.; Bleier, A.; Spence, R.D. [Oak Ridge National Lab., TN (United States)

1996-05-06T23:59:59.000Z

351

Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results  

SciTech Connect (OSTI)

This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

Rechard, R.P. [ed.

1995-03-01T23:59:59.000Z

352

Hanford Site Tank 241-C-108 Residual Waste Contaminant Release Models and Supporting Data  

SciTech Connect (OSTI)

This report presents the results of laboratory characterization, testing, and analysis for a composite sample (designated 20578) of residual waste collected from single-shell tank C-108 during the waste retrieval process after modified sluicing. These studies were completed to characterize concentration and form of contaminant of interest in the residual waste; assess the leachability of contaminants from the solids; and develop release models for contaminants of interest. Because modified sluicing did not achieve 99% removal of the waste, it is expected that additional retrieval processing will take place. As a result, the sample analyzed here is not expected to represent final retrieval sample.

Cantrell, Kirk J.; Krupka, Kenneth M.; Geiszler, Keith N.; Arey, Bruce W.; Schaef, Herbert T.

2010-06-18T23:59:59.000Z

353

Hot isostatic press waste option study report  

SciTech Connect (OSTI)

A Settlement Agreement between the Department of Energy and the State of Idaho mandates that all high-level radioactive waste now stored at the Idaho Chemical Processing Plant be treated so that it is ready to move out of Idaho for disposal by the target date of 2035. This study investigates the immobilization of all Idaho Chemical Processing Plant calcine, including calcined sodium bearing waste, via the process known as hot isostatic press, which produces compact solid waste forms by means of high temperature and pressure (1,050 C and 20,000 psi), as the treatment method for complying with the settlement agreement. The final waste product would be contained in stainless-steel canisters, the same type used at the Savannah River Site for vitrified waste, and stored at the Idaho National Engineering and Environmental Laboratory until a national geological repository becomes available for its disposal. The waste processing period is from 2013 through 2032, and disposal at the High Level Waste repository will probably begin sometime after 2065.

Russell, N.E.; Taylor, D.D.

1998-02-01T23:59:59.000Z

354

Form 200 | Open Energy Information  

Open Energy Info (EERE)

200Legal Abstract Form 200: ApplicationReport for Waste Discharge, current through August 14, 2014. Published NA Year Signed or Took Effect 1997 Legal Citation Form 200:...

355

Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste  

DOE Patents [OSTI]

A method for immobilizing waste chloride salts containing radionuclides such as cesium and strontium and hazardous materials such as barium. A sodalite intermediate is prepared by mixing appropriate amounts of silica, alumina and sodium hydroxide with respect to sodalite and heating the mixture to form the sodalite intermediate and water. Heating is continued to drive off the water to form a water-free intermediate. The water-free intermediate is mixed with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides and hazardous material. The waste salt-intermediate mixture is then compacted and heated under conditions of heat and pressure to form sodalite with the waste salt, radionuclides and hazardous material trapped within the sodalite cage structure. This provides a final product having excellent leach resistant capabilities.

Koyama, Tadafumi (Tokyo, JP)

1994-01-01T23:59:59.000Z

356

Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste  

DOE Patents [OSTI]

This report describes a method for immobilizing waste chloride salts containing radionuclides such as cesium and strontium and hazardous materials such as barium. A sodalite intermediate is prepared by mixing appropriate amounts of silica, alumina and sodium hydroxide with respect to sodalite and heating the mixture to form the sodalite intermediate and water. Heating is continued to drive off the water to form a water-free intermediate. The water-free intermediate is mixed with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides and hazardous material. The waste salt-intermediate mixture is then compacted and heated under conditions of heat and pressure to form sodalite with the waste salt, radionuclides and hazardous material trapped within the sodalite cage structure. This provides a final product having excellent leach resistant capabilities.

Koyama, T.

1992-01-01T23:59:59.000Z

357

Microsoft Word - Sludge Test Area CX Determination Form12172012  

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

Sludge Test Facility at the Transuranic (TRU) Waste Processing Center (TWPC) [CX-TWPC-13-0001] Sludge Test Facility at the Transuranic (TRU) Waste Processing Center (TWPC) [CX-TWPC-13-0001] Program or Field Office: Environmental Management - Oak Ridge Location(s) (City/County/State): Oak Ridge, Tennessee Proposed Action Description: The proposed action is to construct and operate a sludge test facility at the Transuranic (TRU) Waste Processing Center (TWPC) to conduct testing activities for sludge mobilization, mixing, and removal from the Melton Valley Storage Tanks (MVST). The testing is needed to develop appropriate, compliant treatment to a final waste form that will meet the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC). This testing is needed for the mobilization, removal, and treatment of the sludge regardless of

358

Norcal Prototype LNG Truck Fleet: Final Results  

SciTech Connect (OSTI)

U.S. DOE and National Renewable Energy Laboratory evaluated Norcal Waste Systems liquefied natural gas (LNG) waste transfer trucks. Trucks had prototype Cummins Westport ISXG engines. Report gives final evaluation results.

Not Available

2004-07-01T23:59:59.000Z

359

Glass melter system technologies for vitrification of high-sodium-content low-level, radioactive, liquid wastes: Phase 1, SBS demonstration with simulated low-level waste. Final test report  

SciTech Connect (OSTI)

The attached vendor report was prepared for Westinghouse Hanford Company by Babcock & Wilcox as documentation of the Phase I Final Test Report, Cyclone Combustion Melter Demonstration.

Holmes, M.J.; Scotto, M.V.; Shiao, S.Y. [Babcock & Wilcox, Alliance, OH (United States) Research Center

1995-12-31T23:59:59.000Z

360

The second iteration of the Systems Prioritization Method: A systems prioritization and decision-aiding tool for the Waste Isolation Pilot Plant: Volume 3, Analysis for final programmatic recommendations  

SciTech Connect (OSTI)

Systems Prioritization Method (SPM) is a decision-aiding tool developed by Sandia National Laboratories for the US DOE Carlsbad Area Office (DOE/CAO). This tool provides an analytical basis for programmatic decision making for the Waste Isolation Pilot Plant (WIPP). SPM integrates decision-analysis techniques, performance,a nd risk-assessment tools, and advanced information technology. Potential outcomes of proposed activities and combination of activities are used to calculate a probability of demonstrating compliance (PDC) with selected regulations. The results are presented in a decision matrix showing cost, duration, and maximum PDC for all activities in a given cost and duration category. This is the third and final volume in the series which presents the analysis for final programmatic recommendations.

Prindle, N.H.; Boak, D.M.; Weiner, R.F. [and others] [and others

1996-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "final waste form" 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

The role of NDE in nuclear waste clean-up  

SciTech Connect (OSTI)

With the end of the cold war, the need for large production capability of nuclear weapons has diminished; but concerns have arisen about the condition of the waste and the waste storage tanks from weapon production. The major concern is the potential for the waste to somehow contaminate the water, soil, and air around a plant. To completely solve the problem associated with the nuclear waste, the waste must be ultimately retrieved from the tanks, treated, separated into low-level and high-level waste streams, and then put into a final disposal form that will encapsulate the waste so that it will not penetrate the environment. Going through the process of safely retrieving, characterizing, treating, and disposing of the waste requires a large amount of nondestructive evaluation (NDE) sensor technology. The purpose of this paper is to provide a brief description of the types of NDE technologies needed to provide a safe and adequate solution to the nuclear-waste storage and disposal issue.

Light, G.M. [Southwest Research Inst., San Antonio, TX (United States)

1995-08-01T23:59:59.000Z

362

Final Report - Development of New Pressure Swing Adsorption (PSA) Technology to Recover High Valued Products from Chemical Plant and Refinery Waste Systems  

SciTech Connect (OSTI)

Project Objective was to extend pressure swing adsorption (PSA) technology into previously under-exploited applications such as polyolefin production vent gas recovery and H2 recovery from refinery waste gases containing significant amounts of heavy hydrocarbons, aromatics, or H2S.

Keith Ludwig

2004-06-14T23:59:59.000Z

363

Assessment of fluidized-bed-combustion solid wastes for land disposal. Volume 5. Appendix G. Final report, June 1980-June 1984  

SciTech Connect (OSTI)

This report, Appendix G, gives results of: (1) an analysis of representative samples of fluidized-bed-combustion (FBC) wastes, (2) an assessment of the characteristics of leachates generated from the wastes under laboratory and field (landfill) conditions, (3) a characterization of the attenuation of the leachates by earth materials that are typical of disposal settings, (4) development of a way to predict the leachate generation behavior of FBC wastes under landfill conditions on the basis of laboratory test results by establishing a rigorous statistical relationship between the laboratory and field-leaching results, and (5) assessment of the compatibility of commonly used landfill-liner materials with FBC waste leachates. FBC is an emerging energy technology that holds promise for both high efficiency of energy conversion and minimization of adverse air-quality impacts. A major advantage of FBC is that high-sulfur coal can be burned without the use of flue-gas desulfurization equipment to meet air-quality standards. The solid residues that are generated in an FBC unit are usually larger in volume and have different properties than the typical bottom ash from a conventional boiler.

Williamson, H.J.; Grimshaw, T.W.; Dunn, J.E.

1985-02-01T23:59:59.000Z

364

Assessment of fluidized-bed-combustion solid wastes for land disposal. Volume 3. Appendices D and E. Final report, June 1980-June 1984  

SciTech Connect (OSTI)

This report, Appendices D and E, gives results of: (1) an analysis of representative samples of fluidized-bed-combustion (FBC) wastes, (2) an assessment of the characteristics of leachates generated from the wastes under laboratory and field (landfill) conditions, (3) a characterization of the attenuation of the leachates by earth materials that are typical of disposal settings, (4) development of a way to predict the leachate generation behavior of FBC wastes under landfill conditions on the basis of laboratory test results by establishing a rigorous statistical relationship between the laboratory and field-leaching results, and (5) assessment of the compatibility of commonly used landfill-liner materials with FBC waste leachates. FBC is an emerging energy technology that holds promise for both high efficiency of energy conversion and minimization of adverse air-quality impacts. A major advantage of FBC is that high-sulfur coal can be burned without the use of flue-gas desulfurization equipment to meet air-quality standards. The solid residues that are generated in an FBC unit are usually larger in volume and have different properties than the typical bottom ash from a conventional boiler.

Minear, R.A.; Grimshaw, T.W.; Little, W.M.

1985-02-01T23:59:59.000Z

365

Assessment of fluidized-bed-combustion solid wastes for land disposal. Volume 6. Appendix H. Final report, June 1980-June 1984  

SciTech Connect (OSTI)

This report, Appendix H, gives results of: (1) an analysis of representative samples of fluidized-bed-combustion (FBC) wastes, (2) an assessment of the characteristics of leachates generated from the wastes underlaboratory and field (landfill) conditions, (3) a characterization of the attenuation of the leachates by earth materials that are typical of disposal settings, (4) development of a way to predict the leachate generation behavior of FBC wastes under landfill conditions on the basis of laboratory test results by establishing a rigorous statistical relationship between the laboratory and field-leaching results, and (5) assessment of the compatibility of commonly used landfill-liner materials with FBC waste leachates. FBC is an emerging energy technology that holds promise for both high efficiency of energy conversion and minimization of adverse air-quality impacts. A major advantage of FBC is that high-sulfur coal can be burned without the use of flue-gas desulfurization equipment to meet air-quality standards. The solid residues that are generated in an FBC unit are usually larger in volume and have different properties than the typical bottom ash from a conventional boiler.

Williamson, H.J.; Heinrich, D.L.; Grimshaw, T.W.

1985-02-01T23:59:59.000Z

366

Assessment of fluidized-bed-combustion solid wastes for land disposal. Volume 7. Appendix I. Final report, June 1980-June 1984  

SciTech Connect (OSTI)

This report, Appendix 1, gives results of: (1) an analysis of representative samples of fluidized-bed-combustion (FBC) wastes, (2) an assessment of the characteristics of leachates generated from the wastes under laboratory and field (landfill) conditions, (3) a characterization of the attenuation of the leachates by earth materials that are typical of disposal settings, (4) development of a way to predict the leachate generation behavior of FBC wastes under landfill conditions on the basis of laboratory test results by establishing a rigorous statistical relationship between the laboratory and field-leaching results, and (5) assessment of the compatibility of commonly used landfill-liner materials with FBC waste leachates. FBC is an emerging energy technology that holds promise for both high efficiency of energy conversion and minimization of adverse air-quality impacts. A major advantage of FBC is that high-sulfur coal can be burned without the use of flue-gas desulfurization equipment to meet air-quality standards. The solid residues that are generated in an FBC unit are usually larger in volume and have different properties than the typical bottom ash from a conventional boiler.

Eklund, A.G.; Grimshaw, T.W.; Minear, R.A.

1985-02-01T23:59:59.000Z

367

Assessment of fluidized-bed-combustion solid wastes for land disposal. Volume 2. Appendices A through C. Final report, June 1980-June 1984  

SciTech Connect (OSTI)

This report Appendices A through C, gives results of: (1) an analysis of representative samples of fluidized-bed-combustion (FBC) wastes, (2) an assessment of the characteristics of leachates generated from the wastes under laboratory and field (landfill) conditions, (3) a characterization of the attenuation of the leachates by earth materials that are typical of disposal settings, (4) development of a way to predict the leachate generation behavior of FBC wastes under landfill conditions on the basis of laboratory test results by establishing a rigorous statistical relationship between the laboratory and field-leaching results, and (5) assessment of the compatibility of commonly used landfill-liner materials with FBC waste leachates. FBC is an emerging energy technology that holds promise for both high efficiency of energy conversion and minimization of adverse air-quality impacts. A major advantage of FBC is that high-sulfur coal can be burned without the use of flue-gas desulfurization equipment to meet air-quality standards. The solid residues that are generated in an FBC unit are usually larger in volume and have different properties than the typical bottom ash from a conventional boiler.

Little, W.M.; Gibson, T.S.; Grimshaw, T.W.; Eklund, A.G.

1985-02-01T23:59:59.000Z

368

Assessment of fluidized-bed-combustion solid wastes for land disposal. Volume 1. Final report. Report for June 1980-June 1984  

SciTech Connect (OSTI)

This report, including 10 appendices, gives results of: (1) an analysis of representative samples of fluidized-bed-combustion (FBC) wastes, (2) an assessment of the characteristics of leachates generated from the wastes under laboratory and field (landfill) conditions, (3) a characterization of the attenuation of the leachates by earth materials that are typical of disposal settings, (4) development of a way to predict the leachate generation behavior of FBC wastes under landfill conditions on the basis of laboratory test results by establishing a rigorous statistical relationship between the laboratory and field-leaching results, and (5) assessment of the compatibility of commonly used landfill-liner materials with FBC waste leachates. FBC is an emerging energy technology that holds promise for both high efficiency of energy conversion and minimization of adverse air-quality impacts. A major advantage of FBC is that high-sulfur coal can be burned without the use of flue-gas desulfurization equipment to meet air-quality standards. The solid residues that are generated in an FBC unit are usually larger in volume and have different properties than the typical bottom ash from a conventional boiler.

Grimshaw, T.W.; Minear, R.A.; Eklund, A.G.; Little, W.M.; Dunn, J.E.

1985-02-01T23:59:59.000Z

369

Assessment of fluidized-bed-combustion solid wastes for land disposal. Volume 4. Appendix F. Final report, June 1980-June 1984  

SciTech Connect (OSTI)

This report, Appendix F, gives results of: (1) an analysis of representative samples of fluidized-bed-combustion (FBC) wastes, (2) an assessment of the characteristics of leachates generated from the wastes under laboratory and field (landfill) conditions, (3) a characterization of the attenuation of the leachates by earth materials that are typical of disposal settings, (4) development of a way to predict the leachate generation behavior of FBC wastes under landfill conditions on the basis of laboratory test results by establishing a rigorous statistical relationship between the laboratory and field-leaching results, and (5) assessment of the compatibility of commonly used landfill-liner materials with FBC waste leachates. FBC is an emerging energy technology that holds promise for both high efficiency of energy conversion and minimization of adverse air-quality impacts. A major advantage of FBC is that high-sulfur coal can be burned without the use of flue-gas desulfurization equipment to meet air-quality standards. The solid residues that are generated in an FBC unit are usually larger in volume and have different properties than the typical bottom ash from a conventional boiler.

Eklund, A.G.; Grimshaw, T.W.

1985-02-01T23:59:59.000Z

370

Assessment of fluidized-bed-combustion solid wastes for land disposal. Volume 8. Appendix J. Final report, June 1980-June 1984  

SciTech Connect (OSTI)

This report, Appendix J, gives results of: (1) an analysis of representative samples of fluidized-bed-combustion (FBC) wastes, (2) an assessment of the characteristics of leachates generated from the wastes under laboratory and field (landfill) conditions, (3) a characterization of the attenuation of the leachates by earth materials that are typical of disposal settings, (4) development of a way to predict the leachate generation behavior of FBC wastes under landfill conditions on the basis of laboratory test results by establishing a rigorous statistical relationship between the laboratory and field-leaching results, and (5) assessment of the compatibility of commonly used landfill-liner materials with FBC waste leachates. FBC is an emerging energy technology that holds promise for both high efficiency of energy conversion and minimization of adverse air-quality impacts. A major advantage of FBC is that high-sulfur coal can be burned without the use of flue-gas desulfurization equipment to meet air-quality standards. The solid residues that are generated in an FBC unit are usually larger in volume and have different properties than the typical bottom ash from a conventional boiler.

Williamson, H.J.; South, R.C.; Grimshaw, T.W.

1985-02-01T23:59:59.000Z

371

EIS-0026-S2: Final Supplemental Environmental Impact Statement...  

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

S2: Final Supplemental Environmental Impact Statement EIS-0026-S2: Final Supplemental Environmental Impact Statement Carlsbad Area Office, Waste Isolation Pilot Plant Disposal...

372

High level nuclear waste  

SciTech Connect (OSTI)

The DOE Division of Waste Products through a lead office at Savannah River is developing a program to immobilize all US high-level nuclear waste for terminal disposal. DOE high-level wastes include those at the Hanford Plant, the Idaho Chemical Processing Plant, and the Savannah River Plant. Commercial high-level wastes, for which DOE is also developing immobilization technology, include those at the Nuclear Fuel Services Plant and any future commercial fuels reprocessing plants. The first immobilization plant is to be the Defense Waste Processing Facility at Savannah River, scheduled for 1983 project submission to Congress and 1989 operation. Waste forms are still being selected for this plant. Borosilicate glass is currently the reference form, but alternate candidates include concretes, calcines, other glasses, ceramics, and matrix forms.

Crandall, J L

1980-01-01T23:59:59.000Z

373

TRANSCOM (FINAL)  

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

radioactive materials and to maintain communication with the drivers as a public safeguard. Trucks that transport transuranic waste to the Waste Isolation Pilot Plant (WIPP)...

374

Remaining Sites Verification Package for the 100-F-31, 144-F Sanitary Sewer System, Waste Site Reclassification Form 2006-033  

SciTech Connect (OSTI)

The 100-F-31 waste site is a former septic system that supported the inhalation laboratories, also referred to as the 144-F Particle Exposure Laboratory (132-F-2 waste site), which housed animals exposed to particulate material. The 100-F-31 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2006-08-24T23:59:59.000Z

375

Zero Waste, Renewable Energy & Environmental  

E-Print Network [OSTI]

· Dioxins & Furans · The `State of Waste' in the US · WTE Technologies · Thermal Recycling ­ Turnkey dangerous wastes in the form of gases and ash, often creating entirely new hazards, like dioxins and furans

Columbia University

376

Evaluation of packed-bed and fluidized-bed cell technology for the destruction and removal of contaminants in alkaline waste solutions. Final report  

SciTech Connect (OSTI)

Disposing of the large quantity of nuclear waste that has been produced within the DOE complex is an area of active research and development. Electrochemical processes have been reported for the treatment of alkaline wastes including the destruction of nitrate and nitrite and the removal of metals such as Tc and Ru. Electrolytic recovery of metals from dilute solution has been reported using conventional porous electrodes such as felt electrode and reticulated electrode, but use of such electrodes is limited. The pores of such electrodes become blocked as a consequence of metal deposition. If an attempt is made to regenerate these electrodes by dissolution of the deposited metals, oxygen evolution on the matrix competes with dissolution of metals deposited within the pores. On the other hand, the use of three dimensional packed-bed and fluidized-bed electrodes having large surface area per unit volume would offer an improvement on felt or reticulated system because of the greater ease of regeneration.

Hobbs, D.T. [Westinghouse Savannah River Co., Aiken, SC (United States); Bockris, J.O.M.; Kim, Jinseong [Texas A& M Univ., College Station, TX (United States). Surface Electrochemistry Laboratory

1995-12-31T23:59:59.000Z

377

Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada  

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

Contents Contents CR-iii TABLE OF CONTENTS Section Page 8. Transportation Modes, Routes, Affected Environment, and Impacts............................................ CR8-1 8.1 General Opposition to Transporting Spent Nuclear Fuel and High-Level Radioactive Waste ............................................................................................................ CR8-6 8.2 Number of Shipments ..................................................................................................... CR8-37 8.3 Transportation Modes and Routes .................................................................................. CR8-41 8.3.1 State Highway 127, Hoover Dam, Nevada Department of Transportation Alternatives ..............................................................................................................

378

Radiological health review of the Final Environmental Impact Statement Waste Isolation Pilot Plant. Volumes 1 and 2. DOE/EIS-0026  

SciTech Connect (OSTI)

Purpose of the Environmental Evaluation Group (EEG) is to conduct an independent technical evaluation of the potential radiation exposure to people from the proposed Federal radioactive Waste Isolation Pilot Plant (WIPP) near Carlsbad, in order to protect the public health and safety and ensure that there is minimal environmental degradation. Analyses are conducted of reports issued by the US DOE and its contractors, other Federal agencies and other organizations, as they relate to the potential health, safety and environmental impacts from WIPP.

Not Available

1981-01-01T23:59:59.000Z

379

This is the pre-peer reviewed version of the following article: "Can hydraulic fracturing make Poland self-sufficient in natural gas?", which will be published in final form in a  

E-Print Network [OSTI]

Poland self-sufficient in natural gas?", which will be published in final form in a special issue-4296 Can hydraulic fracturing make Poland self-sufficient in natural gas? Kjell Alekletta,b,* , Tadeusz to be able to replace gas from Russia with domestic natural gas production and eventually to become self

Patzek, Tadeusz W.

380

Modeling of hydrologic conditions and solute movement in processed oil shale waste embankments under simulated climatic conditions. Final report, November 1995  

SciTech Connect (OSTI)

A study is described on the hydrological and geotechnical behavior of an oil shale solid waste. The objective was to obtain information which can be used to assess the environmental impacts of oil shale solid waste disposal in the Green River Basin. The spent shale used in this study was combusted by the Lurgi-Ruhrgas process by Rio Blanco Oil Shale Company, Inc. Laboratory bench-scale testing included index properties, such as grain size distribution and Atterberg limits, and tests for engineering properties including hydraulic conductivity and shear strength. Large-scale tests were conducted on model spent shale waste embankments to evaluate hydrological response, including infiltration, runoff, and seepage. Large-scale tests were conducted at a field site in western Colorado and in the Environmental Simulation Laboratory (ESL)at the University of Wyoming. The ESL tests allowed the investigators to control rainfall and temperature, providing information on the hydrological response of spent shale under simulated severe climatic conditions. All experimental methods, materials, facilities, and instrumentation are described in detail, and results are given and discussed. 34 refs.

NONE

1995-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "final waste form" 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

Pioneering Nuclear Waste Disposal  

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

T h e W a s t e I s o l a t i o n P i l o t P l a n t DOE 1980. Final Environmental Impact Statement, Waste Isolation Pilot Plant. DOE/EIS-0026, Washington, DC, Office of Environmental Management, U.S. Department of Energy. DOE 1981. Waste Isolation Pilot Plant (WIPP): Record of Decision. Federal Register, Vol. 46, No. 18, p. 9162, (46 Federal Register 9162), January 28, 1981. U.S. Department of Energy. DOE 1990. Final Supplement Environmental Impact Statement, Waste Isolation Pilot Plant. DOE/EIS-0026-FS, Washington, DC, Office of Environmental Management, U.S. Department of Energy. DOE 1990. Record of Decision: Waste Isolation Pilot Plant. Federal Register, Vol. 55, No. 121, 25689-25692, U.S. Department of Energy. DOE 1994. Comparative Study of Waste Isolation Pilot Plant (WIPP) Transportation Alternatives.

382

Waste acceptance and waste loading for vitrified Oak Ridge tank waste  

SciTech Connect (OSTI)

The Office of Science and Technology of the DOE has funded a joint project between the Oak Ridge National Laboratory (ORNL) and the Savannah River Technology Center (SRTC) to evaluate vitrification and grouting for the immobilization of sludge from ORNL tank farms. The radioactive waste is from the Gunite and Associated Tanks (GAAT), the Melton Valley Storage Tanks (MVST), the Bethel Valley Evaporator Service Tanks (BVEST), and the Old Hydrofractgure Tanks (OHF). Glass formulation development for sludge from these tanks is discussed in an accompanying article for this conference (Andrews and Workman). The sludges contain transuranic radionuclides at levels which will make the glass waste form (at reasonable waste loadings) TRU. Therefore, one of the objectives for this project was to ensure that the vitrified waste form could be disposed of at the Waste Isolation Pilot Plant (WIPP). In order to accomplish this, the waste form must meet the WIPP Waste Acceptance Criteria (WAC). An alternate pathway is to send the glass waste forms for disposal at the Nevada Test Site (NTS). A sludge waste loading in the feed of 6 wt percent will lead to a waste form which is non-TRU and could potentially be disposed of at NTS. The waste forms would then have to meet the requirements of the NTS WAC. This paper presents SRTC`s efforts at demonstrating that the glass waste form produced as a result of vitrification of ORNL sludge will meet all the criteria of the WIPP WAC or NTS WAC.

Harbour, J.R.; Andrews, M.K.

1997-06-06T23:59:59.000Z

383

EA-1962: Analysis for Below Grade Suspect Transuranic (TRU) Waste...  

Energy Savers [EERE]

of waste type and determination of a final disposal path. Per DOE Order 435.1, Change 1, Radioactive Waste Management, and its associated guide, legacy waste at Los Alamos...

384

SRS - Programs - Waste Solidification  

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

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

385

Assessment of solid-waste characteristics and control technology for oil-shale retorting. Final report for September 1983-February 1985  

SciTech Connect (OSTI)

The report presents information on oil-shale deposits in the eastern and western parts of the United States, their geological subdivisions, locations, tonnage, and physical and chemical characteristics. Characteristics of solid and liquid wastes produced from various oil-shale-processing technologies and control methods are presented. Also included are results from an experimental study to construct liners and covers for disposal of spent shale. A compilation of available data on the auto-ignition potential of raw and spent shales indicates a similarity between raw-shale fines and bituminous coals.

Agarwal, A.K.

1986-05-01T23:59:59.000Z

386

Glass Development for Treatment of LANL Evaporator Bottoms Waste  

SciTech Connect (OSTI)

Vitrification is an attractive treatment option for meeting the stabilization and final disposal requirements of many plutonium (Pu) bearing materials and wastes at the Los Alamos National Laboratory (LANL) TA-55 facility, Rocky Flats Environmental Technology Site (RFETS), Hanford, and other Department of Energy (DOE) sites. The Environmental Protection Agency (EPA) has declared that vitrification is the "best demonstrated available technology" for high- level radioactive wastes (HLW) (Federal Register 1990) and has produced a handbook of vitriilcation technologies for treatment of hazardous and radioactive waste (US EPA, 1992). This technology has been demonstrated to convert Pu-containing materials (Kormanos, 1997) into durable (Lutze, 1988) and accountable (Forsberg, 1995) waste. forms with reduced need for safeguarding (McCulhun, 1996). The composition of the Evaporator Bottoms Waste (EVB) at LANL, like that of many other I%-bearing materials, varies widely and is generally unpredictable. The goal of this study is to optimize the composition of glass for EVB waste at LANL, and present the basic techniques and tools for developing optimized glass compositions for other Pu-bearing materials in the complex. This report outlines an approach for glass formulation with fixed property restrictions, using glass property-composition databases. This approach is applicable to waste glass formulation for many variable waste streams and vitrification technologies.. Also reported are the preliminary property data for simulated evaporator bottom glasses, including glass viscosity and glass leach resistance using the Toxicity Characteristic Leaching Procedure (TCLP).

DE Smith; GF Piepel; GW Veazey; JD Vienna; ML Elliott; RK Nakaoka; RP Thimpke

1998-11-20T23:59:59.000Z

387

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

SciTech Connect (OSTI)

This document provides baseline inventories of transuranic wastes for the WIPP facility. Information on waste forms, forecasting of future inventories, and waste stream originators is also provided. A diskette is provided which contains the inventory database.

NONE

1995-02-01T23:59:59.000Z

388

Remaining Sites Verification Package for the 1607-F1 Sanitary Sewer System (124-F-1) and the 100-F-26:8 (1607-F1) Sanitary Sewer Pipelines Waste Sites, Waste Site Reclassification Form 2005-004  

SciTech Connect (OSTI)

The 100-F-26:8 waste site consisted of the underground pipelines that conveyed sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office to the 1607-F1 septic tank. The site has been remediated and presently exists as an open excavation. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2008-03-14T23:59:59.000Z

389

Synthesizing Optimal Waste Blends  

Science Journals Connector (OSTI)

Vitrification of tank wastes to form glass is a technique that will be used for the disposal of high-level waste at Hanford. ... Durability restrictions ensure that the resultant glass meets the quantitative criteria for disposal/long-term storage in a repository. ... If glasses are formulated to minimize the volume of glass that would be produced, then the cost of processing the waste and storing the resultant glass would be greatly reduced. ...

Venkatesh Narayan; Urmila M. Diwekar; Mark Hoza

1996-10-08T23:59:59.000Z

390

Iron phosphate compositions for containment of hazardous metal waste  

DOE Patents [OSTI]

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

Day, Delbert E. (Rolla, MO)

1998-01-01T23:59:59.000Z

391

Final Technical Report  

SciTech Connect (OSTI)

The project, ?Capital Investment to Fund Equipment Purchases and Facility Modifications to Create a Sustainable Future for EnergyXchange? served to replace landfill gas energy with alternative energy resources, primarily solar and wood waste. This is the final project closeout report.

Gilbert, Chris [Altamont Environmental, Inc.] [Altamont Environmental, Inc.

2014-11-13T23:59:59.000Z

392

Remaining Sites Verification Package for the 116-F-16, PNL Outfall and the 100-F-43, PNL Outfall Spillway, Waste Site Reclassification Form 2006-046  

SciTech Connect (OSTI)

The 100-F-43 waste site is the portion of the former discharge spillway for the PNL Outfall formerly existing above the ordinary high water mark of the Columbia River. The spillway consisted of a concrete flume used to discharge waste effluents from the 100-F Experimental Animal Farm. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2006-09-14T23:59:59.000Z

393

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

394

First annual report RCRA post-closure monitoring and inspections for the U-3fi waste unit. Final report, July 1995--October 1996  

SciTech Connect (OSTI)

This annual Neutron Soil Moisture Monitoring report provides an analysis and summary for site inspections, meteorological information, and neutron soil moisture monitoring data obtained at the U-3fi RCRA Unit, located in Area 3 of the Nevada Site (NTS), Nye County, Nevada during the July 1995 to October 1996 period. Inspections of the U-3fi RCRA Unit are conducted to determine and document the physical condition of the covers, facilities, and any unusual conditions that could impact the proper operation of the waste unit closure. The objective of the neutron logging is to monitor the soil moisture conditions along the 420 ft ER3-3 borehole and detect changes that may be indicative of moisture movement in the regulated interval. This is the first annual report on the U-3fi closure and includes the first year baseline monitoring data as well as one quarter of compliance monitoring data.

Emer, D.F.

1997-01-01T23:59:59.000Z

395

Waste IncIneratIon and Waste PreventIon  

E-Print Network [OSTI]

disposing of waste, it also makes consider- able amounts of energy available in the form of electricity emissions annu- ally. About 50 percent of the energy contained in residual municipal waste comes from- sions from the fossil waste fraction and the fos- sil energy purchased from external sources

396

Sorting and disposal of hazardous laboratory Radioactive waste  

E-Print Network [OSTI]

Sorting and disposal of hazardous laboratory waste Radioactive waste Solid radioactive waste or in a Perspex box. Liquid radioactive waste collect in a screw-cap plastic bottle, ½ or 1 L size. Place bottles in a tray to avoid spill Final disposal of both solid and radioactive waste into the yellow barrel

Maoz, Shahar

397

E-Print Network 3.0 - anaerobic thermophiles final Sample Search...  

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

for treating this waste stream since it results in two valuable final products, biogas and ... Source: Columbia University - Waste-to-Energy Research and Technology Council...

398

Final Supplemental Environmental Impact Statement  

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

Supplemental Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada Summary U.S. Department of Energy Office of Civilian Radioactive Waste Management DOE/EIS-0250F-S1 June 2008 Final Supplemental Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada Summary U.S. Department of Energy Office of Civilian Radioactive Waste Management DOE/EIS-0250F-S1 June 2008 Foreword COVER SHEET RESPONSIBLE AGENCY: U.S. Department of Energy (DOE) TITLE: Final Supplemental Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada

399

Nuclear waste management. Quarterly progress report, January-March 1980  

SciTech Connect (OSTI)

Reported are: high-level waste immobilization, alternative waste forms, nuclear waste materials characterization, TRU waste immobilization, TRU waste decontamination, krypton solidification, thermal outgassing, iodine-129 fixation, unsaturated zone transport, well-logging instrumentation development, mobile organic complexes of fission products, waste management system and safety studies, assessment of effectiveness of geologic isolation systems, waste/rock interactions, engineered barriers, criteria for defining waste isolation, and spent fuel and pool component integrity. (DLC)

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

1980-06-01T23:59:59.000Z

400

EA-0437: Final Environmental Assessment | Department of Energy  

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

437: Final Environmental Assessment 437: Final Environmental Assessment EA-0437: Final Environmental Assessment Process Equipment Waste and Process Waste Liquid Collection Systems Idaho Chemical Processing Plant, Idaho National Engineering Laboratory This Environmental Assessment describes and evaluates the environmental impacts of Idaho National Engineering Laboratory's proposed action (and alternatives to the proposed action). The proposed action, which would be to install new waste lines and associated equipment in the Process Equipment Waste and Process Waste Liquid systems, seeks to bring those existing parts of a waste management facility into compliance with applicable state and federal environmental regulatory requirements (49 CFR 265 Subpart J and Idaho Administrative Procedures Act Title I, Chapter 5,

Note: This page contains sample records for the topic "final waste form" 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

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

SciTech Connect (OSTI)

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

L. M. Dittmer

2008-03-03T23:59:59.000Z

402

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

SciTech Connect (OSTI)

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

L. M. Dittmer

2007-03-21T23:59:59.000Z

403

Remaining Sites Verification Package for the 100-B-20, 1716-B Maintenance Garage Underground Tank, Waste Site Reclassification Form 2006-019  

SciTech Connect (OSTI)

The 100-B-20 waste site, located in the 100-BC-1 Operable Unit of the Hanford Site, consisted of an underground oil tank that once serviced the 1716-B Maintenance Garage. The selected action for the 100-B-20 waste site involved removal of the oil tanks and their contents and demonstrating through confirmatory sampling that all cleanup goals have been met. In accordance with this evaluation, a reclassification status of interim closed out has been determined. The results demonstrate that the site will support future unrestricted land uses that can be represented by a rural-residential scenario. These results also show that residual concentrations support unrestricted future use of shallow zone soil and that contaminant levels remaining in the soil are protective of groundwater and the Columbia River.

L. M. Dittmer

2006-09-27T23:59:59.000Z

404

Remaining Sites Verification Package for the 116-F-8, 1904-F Outfall Structure and the 100-F-42, 1904-F Spillway, Waste Site Reclassification Form 2006-038  

SciTech Connect (OSTI)

The 116-F-8 waste site is the former 1904-F Outfall Structure used to discharge reactor cooling water effluent fro mthe 107-F Retention Basin to the Columbia River. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2006-09-25T23:59:59.000Z

405

DOE FG02-03ER63557: Final Technical Report: Reactivity of Primary Soil Minerals and Secondary Precipitates Beneath Leaking Hanford Waste Tanks  

SciTech Connect (OSTI)

The purpose of the project was to investigate rates and mechanisms of reactions between primary sediment minerals and key components of waste tank solutions that leaked into the subsurface at the Hanford Site. Results were expected to enhance understanding of processes that cause (1) changes in porosity and permeability of the sediment and resultant changes in flow paths of the contaminant plumes, (2) formation of secondary precipitates that can take up contaminants in their structures, and (3) release of mineral components that can drive redox reactions affecting dissolved contaminant mobility. Measured rates can also be used directly in reactive transport models. Project tasks included (1) measurement of the dissolution rates of biotite mica from low to high pH and over a range of temperature relevant to the Hanford subsurface, (2) measurement of dissolution rates of quartz at high pH and in the presence of dissolved alumina, (3) measurement of the dissolution rates of plagioclase feldspar in high pH, high nitrate, high Al-bearing solutions characteristic of the BX tank farms, (4) incorporation of perrhenate in iron-oxide minerals as a function of pH, and (5) initiation of experiments to measure the formation of uranium(VI)-silicate phases under ambient conditions. Task 2 was started under a previous grant from the Environmental Management Science Program and Task 4 was partially supported by a grant to the PI from the Geosciences Program, Office of Basic Energy Sciences. Task 5 was continued under a subsequent grant from the Environmental Remediation Sciences Program, Office of Biological and Environmental Research.

Kathryn L. Nagy

2009-05-04T23:59:59.000Z

406

Final Systems Development Report for the Clark County Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Yucca Mountain, NV  

SciTech Connect (OSTI)

The Systems Development Report represents the third major step in the Clark County Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Yucca Mound Nevada. The first of these steps was to forge a Research Design that would serve as a guide for the overall research process. The second step was the construction of the Base Case, the purpose of which was to describe existing conditions in Clark County in the specified analytic areas of Economic-Demographic/Fiscal, Emergency Planning and Management, Transportation and Sociocultural analysis. The base case description will serve as a basis for assessing changes in these topic areas that might result from the Yucca Mountain project. These changes will be assessed by analyzing conditions with and without repository development in the county. Prior to performing such assessments, however, the snapshot type of data found in the base case must be operationalized or systematized to allow for more dynamic data utilization. In other words, a data system that can be used to analyze the consequences of the introduction of different variables (or variable values) in the Clark County context must be constructed. Such a system must be capable of being updated through subsequent data collection and monitoring efforts to both provide a rolling base case and supply information necessary to construct trend analyses. For example, during the Impact Assessment phase of the study process, the without repository analysis is accomplished by analyzing growth for the county given existing conditions and likely trends. These data are then compared to the with Yucca Mountain project conditions anticipated for the county. Similarly, once the emergency planning management and response needs associated with the repository are described, these needs will be juxtaposed against existing (and various future) capacity(ies) in order to determine the nature and magnitude of impacts in this analytic area. Analogous tasks will be performed for the other analytic areas detailed in the Base Case and outlined below.

NONE

1992-06-18T23:59:59.000Z