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Sample records for mixed waste vitrification

  1. Vitrification development for mixed wastes

    International Nuclear Information System (INIS)

    Merrill, R.; Whittington, K.; Peters, R.

    1995-02-01

    Vitrification is a promising approach to waste-form immobilization. It destroys hazardous organic compounds and produces a durable and highly stable glass. Vitrification tests were performed on three surrogate wastes during fiscal year 1994; 183-H Solar Evaporation Basin waste from Hanford, bottom ash from the Oak Ridge TSCA incinerator, and saltcrete from Rocky Flats. Preliminary glass development involved melting trials followed by visual homogeneity examination, short-duration leach tests on glass specimens, and long-term leach tests on selected glasses. Viscosity and electrical conductivity measurements were taken for the most durable glass formulations. Results for the saltcrete are presented in this paper and demonstrate the applicability of vitrification technology to this mixed waste

  2. Transportable Vitrification System Demonstration on Mixed Waste

    International Nuclear Information System (INIS)

    Zamecnik, J.R.; Whitehouse, J.C.; Wilson, C.N.; Van Ryn, F.R.

    1998-01-01

    This paper describes preliminary results from the first demonstration of the Transportable Vitrification System (TVS) on actual mixed waste. The TVS is a fully integrated, transportable system for the treatment of mixed and low-level radioactive wastes. The demonstration was conducted at Oak Ridge's East Tennessee Technology Park (ETTP), formerly known as the K-25 site. The purpose of the demonstration was to show that mixed wastes could be vitrified safely on a 'field' scale using joule-heated melter technology and obtain information on system performance, waste form durability, air emissions, and costs

  3. Mixed Wastes Vitrification by Transferred Plasma

    International Nuclear Information System (INIS)

    Tapia-Fabela, J.; Pacheco-Pacheco, M.; Pacheco-Sotelo, J.; Torres-Reyes, C.; Valdivia-Barrientos, R.; Benitez-Read, J.; Lopez-Callejas, R.; Ramos-Flores, F.; Boshle, S.; Zissis, G.

    2007-01-01

    Thermal plasma technology provides a stable and long term treatment of mixed wastes through vitrification processes. In this work, a transferred plasma system was realized to vitrify mixed wastes, taking advantage of its high power density, enthalpy and chemical reactivity as well as its rapid quenching and high operation temperatures. To characterize the plasma discharge, a temperature diagnostic is realized by means of optical emission spectroscopy (OES). To typify the morphological structure of the wastes samples, scanning electron microscopy (SEM), and X-ray diffraction (XRD) techniques were applied before and after the plasma treatment

  4. Prospects for vitrification of mixed wastes at ANL-E

    International Nuclear Information System (INIS)

    Mazer, J.; No, Hyo.

    1993-01-01

    This report summarizes a study evaluating the prospects for vitrification of some of the mixed wastes at ANL-E. This project can be justified on the following basis: Some of ANL-E's mixed waste streams will be stabilized such that they can be treated as a low-level radioactive waste. The expected volume reduction that results during vitrification will significantly reduce the overall waste volume requiring disposal. Mixed-waste disposal options currently used by ANL-E may not be permissible in the near future without treatment technologies such as vitrification

  5. Transportable Vitrification System: Operational experience gained during vitrification of simulated mixed waste

    International Nuclear Information System (INIS)

    Whitehouse, J.C.; Burket, P.R.; Crowley, D.A.; Hansen, E.K.; Jantzen, C.M.; Smith, M.E.; Singer, R.P.; Young, S.R.; Zamecnik, J.R.; Overcamp, T.J.; Pence, I.W. Jr.

    1996-01-01

    The Transportable Vitrification System (TVS) is a large-scale, fully-integrated, transportable, vitrification system for the treatment of low-level nuclear and mixed wastes in the form of sludges, soils, incinerator ash, and similar waste streams. The TVS was built to demonstrate the vitrification of actual mixed waste at U. S. Department of Energy (DOE) sites. Currently, Westinghouse Savannah River Company (WSRC) is working with Lockheed Martin Energy Systems (LMES) to apply field scale vitrification to actual mixed waste at Oak Ridge Reservation's (ORR) K-25 Site. Prior to the application of the TVS to actual mixed waste it was tested on simulated K-25 B and C Pond waste at Clemson University. This paper describes the results of that testing and preparations for the demonstration on actual mixed waste

  6. Status of vitrification for DOE low-level mixed waste

    International Nuclear Information System (INIS)

    Schumacher, R.F.; Jantzen, C.M.; Plodinec, M.J.

    1993-04-01

    Vitrification is being considered by the Department of Energy for solidification of many low-level mixed waste streams. Some of the advantages, requirements, and potential problem areas are described. Recommendations for future efforts are presented

  7. Vitrification of low-level and mixed wastes

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  8. Vitrification development plan for US Department of Energy mixed wastes

    International Nuclear Information System (INIS)

    Peters, R.; Lucerna, J.; Plodinec, M.J.

    1993-10-01

    This document is a general plan for conducting vitrification development for application to mixed wastes owned by the US Department of Energy. The emphasis is a description and discussion of the data needs to proceed through various stages of development. These stages are (1) screening at a waste site to determine which streams should be vitrified, (2) waste characterization and analysis, (3) waste form development and treatability studies, (4) process engineering development, (5) flowsheet and technical specifications for treatment processes, and (6) integrated pilot-scale demonstration. Appendices provide sample test plans for various stages of the vitrification development process. This plan is directed at thermal treatments which produce waste glass. However, the study is still applicable to the broader realm of thermal treatment since it deals with issues such as off-gas characterization and waste characterization that are not necessarily specific to vitrification. The purpose is to provide those exploring or considering vitrification with information concerning the kinds of data that are needed, the way the data are obtained, and the way the data are used. This will provide guidance to those who need to prioritize data needs to fit schedules and budgets. Knowledge of data needs also permits managers and planners to estimate resource requirements for vitrification development

  9. Vitrification of hazardous and mixed wastes

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Pickett, J.B.; Ramsey, W.G.

    1992-01-01

    Solidification of hazardous/mixed wastes into glass is being examined at the Savannah River Site. The first hazardous/mixed wastes glassified at SRS have been (1) incinerator and (2) nickel plating line (F006) wastes. Solidification of incinerator blowdown and mixtures of incinerator blowdown and incinerator bottom kiln ash have been achieved in Soda (Na 2 O) - Lime (CaO) - Silica (SiO 2 ) glass (SLS) at waste loadings of up to 50 wt%. Solidification of nickel-plating line waste sludges containing depleted uranium have also been achieved in both SLS and borosilicate glasses at waste loadings of 75 wt%. This corresponds to volume reductions of 97% and 81%, respectively. Further studies will examine glassification of: ion exchange zeolites, inorganic filter media, asbestos, glass fiber filters, contaminated soil, cementitious, or other materials in need of remediation

  10. Transportable vitrification system demonstration on mixed waste. Revision 1

    International Nuclear Information System (INIS)

    Zamecnik, J.R.; Whitehouse, J.C.; Wilson, C.N.; Van Ryn, F.R.

    1998-01-01

    The Transportable Vitrification System (TVS) is a large scale, fully integrated, vitrification system for the treatment of low-level and mixed wastes in the form of sludges, soils, incinerator ash, and many other waste streams. It was demonstrated on surrogate waste at Clemson University and at the Oak Ridge Reservation (ORR) prior to treating actual mixed waste. Treatment of a combination of dried B and C Pond sludge and CNF sludge was successfully demonstrated at ORR in 1997. The demonstration produced 7,616 kg of glass from 7,328 kg of mixed wastes with a 60% reduction in volume. Glass formulations for the wastes treated were developed using a combination of laboratory crucible studies with the actual wastes and small melter studies at Clemson with both surrogate and actual wastes. Initial characterization of the B and C Pond sludge had not shown the presence of carbon or fluoride, which required a modified glass formulation be developed to maintain proper glass redox and viscosity. The CNF sludge challenges the glass formulations due to high levels of phosphate and iron. The demonstration was delayed several times by permitting problems, a glass leak, and electrical problems. The demonstration showed that the two wastes could be successfully vitrified, although the design glass production rate was not achieved. The glass produced met the Universal Treatment Standards and the emissions from the TVS were well within the allowable permit limits

  11. Transportable vitrification system demonstration on mixed waste. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J.R.; Whitehouse, J.C. [Westinghouse Savannah River Co., Aiken, SC (United States); Wilson, C.N. [Lockheed Martin Hanford Corp., Richland, WA (United States); Van Ryn, F.R. [Bechtel Jacobs Co., Oak Ridge, TN (United States)

    1998-04-22

    The Transportable Vitrification System (TVS) is a large scale, fully integrated, vitrification system for the treatment of low-level and mixed wastes in the form of sludges, soils, incinerator ash, and many other waste streams. It was demonstrated on surrogate waste at Clemson University and at the Oak Ridge Reservation (ORR) prior to treating actual mixed waste. Treatment of a combination of dried B and C Pond sludge and CNF sludge was successfully demonstrated at ORR in 1997. The demonstration produced 7,616 kg of glass from 7,328 kg of mixed wastes with a 60% reduction in volume. Glass formulations for the wastes treated were developed using a combination of laboratory crucible studies with the actual wastes and small melter studies at Clemson with both surrogate and actual wastes. Initial characterization of the B and C Pond sludge had not shown the presence of carbon or fluoride, which required a modified glass formulation be developed to maintain proper glass redox and viscosity. The CNF sludge challenges the glass formulations due to high levels of phosphate and iron. The demonstration was delayed several times by permitting problems, a glass leak, and electrical problems. The demonstration showed that the two wastes could be successfully vitrified, although the design glass production rate was not achieved. The glass produced met the Universal Treatment Standards and the emissions from the TVS were well within the allowable permit limits.

  12. Vitrification of low level and mixed (radioactive and hazardous) wastes: Lessons learned from high level waste vitrification

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1994-01-01

    Borosilicate glasses will be used in the USA and in Europe immobilize radioactive high level liquid wastes (HLLW) for ultimate geologic disposal. Simultaneously, tehnologies are being developed by the US Department of Energy's (DOE) Nuclear Facility sites to immobilize low-level and mixed (radioactive and hazardous) wastes (LLMW) in durable glass formulations for permanent disposal or long-term storage. Vitrification of LLMW achieves large volume reductions (86--97 %) which minimize the associated long-term storage costs. Vitrification of LLMW also ensures that mixed wastes are stabilized to the highest level reasonably possible, e.g. equivalent to HLLW, in order to meet both current and future regulatory waste disposal specifications The tehnologies being developed for vitrification of LLMW rely heavily on the technologies developed for HLLW and the lessons learned about process and product control

  13. In situ vitrification of mixed wastes: Progress and regulatory status

    International Nuclear Information System (INIS)

    Kindle, C.H.; Barich, J.J. III

    1991-08-01

    In situ vitrification (ISV) technology targets mixed wastes in in situ near-surface environments. Federal laws governing toxic substances (TSCA), hazardous waste (RCRA), and abandoned sites (Superfund) create the need for remediation technology and define the required performance characteristics. The need for ISV depends, in part, on the extent of regulation and how well ISV's demonstrated performance characteristics match up with regulatory criteria. The regulatory requirements are easier to identify and meet in short-duration site- and situation-specific applications of the technology than they are simpler in long-term, generalized applications. ISV's ability to treat both inorganics and organics in a single process supports applications for mixed, hazardous, and radioactive sites of moderate depth (20 ft). The durability of the ISV waste form is a major advantage of the technology when demonstrating permanence of a waste management strategy. Achieving depth and vapor containment assurance are issues being addressed as the ISV process is refined for new applications having different processing concerns. Refinements include moveable electrodes and sheet steel as the material for the containment structure. 16 refs., 4 figs., 6 tabs

  14. Choosing solidification or vitrification for low-level radioactive and mixed waste treatment

    International Nuclear Information System (INIS)

    Gimpel, R.F.

    1992-01-01

    Solidification (making concrete) and vitrification (making glass) are frequently the treatment methods recommended for treating inorganic or radioactive wastes. Solidification is generally perceived as the most economical treatment method. Whereas, vitrification is considered (by many) as the most effective of all treatment methods. Unfortunately, vitrification has acquired the stigma that it is too expensive to receive further consideration as an alternative to solidification in high volume treatment applications. Ironically, economic studies, as presented in this paper, show that vitrification may be more competitive in some high volume applications. Ex-situ solidification and vitrification are the competing methods for treating in excess of 450 000m 3 of low-level radioactive and mixed waste at the Fernald Environmental Management Project (FEMP or simply, Fernald) located near Cincinnati, Ohio. This paper summarizes how Fernald is choosing between solidification and vitrification as the primary waste treatment method

  15. Choosing solidification or vitrification for low-level radioactive and mixed waste treatment

    International Nuclear Information System (INIS)

    Gimpel, R.F.

    1992-01-01

    Solidification (making concrete) and vitrification (making glass) are frequently the treatment methods recommended for treating inorganic or radioactive wastes. Solidification is generally perceived as the most economical treatment method. Whereas, vitrification is considered (by many) as the most effective of all treatment methods. Unfortunately, vitrification has acquired the stigma that it is too expensive to receive further consideration as an alternative to solidification in high volume treatment applications. Ironically, economic studies, as presented in this paper, show that vitrification may be more competitive in some high volume applications. Ex-situ solidification and vitrification are the competing methods for treating in excess of 450,000 m 3 of low-level radioactive and mixed waste at the Fernald Environmental Management Project (FEMP or simply, Fernald) located near Cincinnati, Ohio. This paper summarized how Fernald is choosing between solidification and vitrification as the primary waste treatment method

  16. Vitrification Studies with DOE Low-Level Mixed Waste Wastewater Treatment Sludges

    International Nuclear Information System (INIS)

    Cicero, C.A.; Andrews, M.K.; Bickford, D.F.; Hewlett, K.J.; Bennert, D.M.; Overcamp, T.J.

    1995-01-01

    Vitrification studies with simulated Low Level Mixed Waste (LLMW) sludges were performed at the Savannah River Technology Center (SRTC). These studies focused on finding the optimum glass compositions for four simulated LLMW wastewater treatment sludges and were based on both crucible-scale and pilot-scale studies. Optimum compositions were determined based on the maximum waste loading achievable without sacrificing glass integrity

  17. Pilot scale vitrification studies on hazardous and mixed wastes

    International Nuclear Information System (INIS)

    Bennert, D.M.; Overcamp, T.J.; Compton, K.L.; Sargent, T.N. Jr.; Resce, J.L.

    1993-01-01

    Over the past 30 years, the Department of Energy has committed extensive resources to the development of technologies suitable for the stabilization of high level radioactive waste. The objective of this work is to produce a vitreous wasteform capable of retaining the radioactive fractions in a leach resistant form. In an effort to further the development of technologies based within the DOE Complex, the DOE is making efforts to promote technical transfer initiatives that will bring these technologies to the private sector. To this end, the Department of Energy through the Savannah River Site is working with Clemson University's Environmental Systems Engineering Department to establish a laboratory dedicated to vitrification research. The laboratory is part of a cooperative effort between Westinghouse Savannah River Company, Clemson University, and their industrial partners EnVitCo, Inc., and Stir Melter, Inc

  18. High temperature vitrification of surrogate Savannah River Site (SRS) mixed waste materials

    International Nuclear Information System (INIS)

    Applewhite-Ramsey, A.; Schumacher, R.F.; Spatz, T.L.; Newsom, R.A.; Circeo, L.J.; Danjaji, M.B.

    1995-01-01

    The Savannah River Technology Center (SRTC) has been funded through the DOE Office of Technology Development (DOE-OTD) to investigate high-temperature vitrification technologies for the treatment of diverse low-level and mixed wastes. High temperature vitrification is a likely candidate for processing heterogeneous solid wastes containing low levels of activity. Many SRS wastes fit into this category. Plasma torch technology is one high temperature vitrification method. A trial demonstration of plasma torch processing is being performed at the Georgia Institute of Technology on surrogate SRS wastes. This effort is in cooperation with the Engineering Research and Development Association of Georgia Universities (ERDA) program. The results of phase 1 of these plasma torch trials will be presented

  19. In situ vitrification of a mixed radioactive and hazardous waste site

    International Nuclear Information System (INIS)

    Campbell, B.E.; Koegler, S.S.

    1990-11-01

    A large-scale test of the in situ vitrification (ISV) process was performed on a mixed radioactive and hazardous-chemical contaminated waste site on the Hanford Site in southeastern Washington State. A mixed-waste site was selected for this large-scale test to demonstrate the applicability of ISV to mixed wastes common to many US Department of Energy (DOE) sites. In situ vitrification is a thermal process that converts contaminated soil into a durable, leach-resistant product. Electrodes are inserted into the ground. The goals of the test are to demonstrate at least 99% retention of fission products and hazardous metals in the ISV glass during the test; demonstrate the ability of the ISV off-gas treatment system to process a waste site containing significant quantities of combustible material and demonstrate the ability of ISV to vitrify the site to a depth of 20 ft or greater. The test was completed in April 1990. 5 figs

  20. Reactive Additive Stabilization Process (RASP) for hazardous and mixed waste vitrification

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Pickett, J.B.; Ramsey, W.G.

    1993-01-01

    Solidification of hazardous/mixed wastes into glass is being examined at the Savannah River Site (SRS) for (1) nickel plating line (F006) sludges and (2) incinerator wastes. Vitrification of these wastes using high surface area additives, the Reactive Additive Stabilization Process (RASP), has been determined to greatly enhance the dissolution and retention of hazardous, mixed, and heavy metal species in glass. RASP lowers melt temperatures (typically 1050-- 1150 degrees C), thereby minimizing volatility concerns during vitrification. RASP maximizes waste loading (typically 50--75 wt% on a dry oxide basis) by taking advantage of the glass forming potential of the waste. RASP vitrification thereby minimizes waste disposal volume (typically 86--97 vol. %), and maximizes cost savings. Solidification of the F006 plating line sludges containing depleted uranium has been achieved in both soda-lime-silica (SLS) and borosilicate glasses at 1150 degrees C up to waste loadings of 75 wt%. Solidification of incinerator blowdown and mixtures of incinerator blowdown and bottom kiln ash have been achieved in SLS glass at 1150 degrees C up to waste loadings of 50% using RASP. These waste loadings correspond to volume reductions of 86 and 94 volume %, respectively, with large associated savings in storage costs

  1. The Treatment of Mixed Waste with GeoMelt In-Container Vitrification

    International Nuclear Information System (INIS)

    Finucane, K.G.; Campbell, B.E.

    2006-01-01

    AMEC's GeoMelt R In-Container Vitrification (ICV) TM has been used to treat diverse types of mixed low-level radioactive waste. ICV is effective in the treatment of mixed wastes containing polychlorinated biphenyls (PCBs) and other semi-volatile organic compounds, volatile organic compounds (VOCs) and heavy metals. The GeoMelt vitrification process destroys organic compounds and immobilizes metals and radionuclides in an extremely durable glass waste form. The process is flexible allowing for treatment of aqueous, oily, and solid mixed waste, including contaminated soil. In 2004, ICV was used to treat mixed radioactive waste sludge containing PCBs generated from a commercial cleanup project regulated by the Toxic Substances Control Act (TSCA), and to treat contaminated soil from Rocky Flats Environmental Technology Site. The Rocky Flats soil contained cadmium, PCBs, and depleted uranium. In 2005, AMEC completed a treatability demonstration of the ICV technology on Mock High Explosive from Sandia National Laboratories. This paper summarizes results from these mixed waste treatment projects. (authors)

  2. Plasma Hearth Process vitrification of DOE low-level mixed waste

    International Nuclear Information System (INIS)

    Gillins, R.L.; Geimer, R.M.

    1995-01-01

    The Plasma Hearth Process (PHP) demonstration project is one of the key technology projects in the Department of Energy (DOE) Office of Technology Development Mixed Waste Focus Area. The PHP is recognized as one of the more promising solutions to DOE's mixed waste treatment needs, with potential application in the treatment of a wide variety of DOE mixed wastes. The PHP is a high temperature vitrification process using a plasma arc torch in a stationary, refractory lined chamber that destroys organics and stabilizes the residuals in a nonleaching, vitrified waste form. This technology will be equally applicable to low-level mixed wastes generated by nuclear utilities. The final waste form will be volume reduced to the maximum extent practical, because all organics will have been destroyed and the inorganics will be in a high-density, low void-space form and little or no volume-increasing glass makers will have been added. Low volume and high integrity waste forms result in low disposal costs. This project is structured to ensure that the plasma technology can be successfully employed in radioactive service. The PHP technology will be developed into a production system through a sequence of tests on several test units, both non-radioactive and radioactive. As the final step, a prototype PHP system will be constructed for full-scale radioactive waste treatment demonstration

  3. In situ vitrification of a mixed radioactive and hazardous waste site

    International Nuclear Information System (INIS)

    Koegler, S.S.

    1990-01-01

    This paper reports on a large-scale test of the in situ vitrification (ISV) process being performed on a mixed radioactive and hazardous-chemical contaminated waste site on the Hanford Site in southeastern Washington state. A mixed-waste site was selected for this large-scale test to demonstrate the applicability of ISV to mixed wastes common to many U.S. Department of Energy (DOE) sites. In situ vitrification is a thermal process that converts contaminated soil into a durable, leach-resistant product. Electrodes are inserted into the ground to the desired treatment depth, and a layer of electrically conductive material (a starter path) is placed between the electrodes. Electrical power is applied to the electrodes causing the conductive material to melt, thus melting the surrounding soil. Electrical energy is transferred to the molten soil through Joule (resistance) heating and the soil continues to melt to the desired depth, at which time the power to the electrodes is discontinued. A hood placed over the area to be vitrified allows the off gases from the process to be treated before their release to the atmosphere. After completion of the melt, the molten-soil cools and solidifies, and soil is backfilled over the subsided area

  4. Vitrification of NORM wastes

    International Nuclear Information System (INIS)

    Chapman, C.

    1994-05-01

    Vitrification of wastes is a relatively new application of none of man's oldest manufacturing processes. During the past 25 years it has been developed and accepted internationally for immobilizing the most highly radioactive wastes from spent nuclear fuel. By the year 2005, there will be nine operating high-level radioactive vitrification plants. Many of the technical ''lessons learned'' from this international program can be applied to much less hazardous materials such as naturally occurring radioactive material (NORM). With the deployment of low capital and operating cost systems, vitrification should become a broadly applied process for treating a large variety of wastes. In many situations, the wastes can be transformed into marketable products. This paper will present a general description of waste vitrification, summarize some of its key advantages, provide some test data for a small sample of one NORM, and suggest how this process may be applied to NORM

  5. First Commercial US Mixed Waste Vitrification Facility: Permits, Readiness Reviews, and Delisting of Final Wasteform

    International Nuclear Information System (INIS)

    Pickett, J.B.; Norford, S.W.; Diener, G.A.

    1998-01-01

    Westinghouse Savannah River Co. (WSRC) contracted GTS Duratek (Duratek) to construct and operate the first commercial vitrification facility to treat an F-006 mixed (radioactive/hazardous) waste in the United States. The permits were prepared and submitted to the South Carolina state regulators by WSRC - based on a detailed design by Duratek. Readiness Assessments were conducted by WSRC and Duratek at each major phase of the operation (sludge transfer, construction, cold and radioactive operations, and a major restart) and approved by the Savannah River Department of Energy prior to proceeding. WSRC prepared the first 'Upfront Delisting' petition for a vitrified mixed waste. Lessons learned with respect to the permit strategy, operational assessments, and delisting from this 'privatization' project will be discussed

  6. Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Larson, D.E.; Allen, C.R.; Kruger, O.L.; Weber, E.T.

    1991-10-01

    The Hanford Waste Vitrification Plant (HWVP) is being designed to immobilize pretreated Hanford high-level waste and transuranic waste in borosilicate glass contained in stainless steel canisters. Testing is being conducted in the HWVP Technology Development Project to ensure that adapted technologies are applicable to the candidate Hanford wastes and to generate information for waste form qualification. Empirical modeling is being conducted to define a glass composition range consistent with process and waste form qualification requirements. Laboratory studies are conducted to determine process stream properties, characterize the redox chemistry of the melter feed as a basis for controlling melt foaming and evaluate zeolite sorption materials for process waste treatment. Pilot-scale tests have been performed with simulated melter feed to access filtration for solids removal from process wastes, evaluate vitrification process performance and assess offgas equipment performance. Process equipment construction materials are being selected based on literature review, corrosion testing, and performance in pilot-scale testing. 3 figs., 6 tabs

  7. Superconducting Open-Gradient Magnetic Separation for the Pretreatment of Radioactive or Mixed Waste Vitrification Feeds

    International Nuclear Information System (INIS)

    Nunez', L.; Kaminsky', M.D.; Crawford, C.; Ritter, J.A.

    1999-01-01

    An open-gradient magnetic separation (OGMS) process is being considered to separate deleterious elements from radioactive and mixed waste streams prior to vitrification or stabilization. By physically segregating solid wastes and slurries based on the magnetic properties of the solid constituents, this potentially low-cost process may serve the U.S. Department of Energy (DOE) by reducing the large quantities of glass produced from defense-related high-level waste (HLW). Furthermore, the separation of deleterious elements from low-level waste (LLW) also can reduce the total quantity of waste produced in LLW immobilization activities. Many HLW 'and LLW waste' streams at both Hanford and the Savannah River Site (SRS) include constituents deleterious to the durability of borosilicate glass and the melter many of the constituents also possess paramagnetism. For example, Fe, Cr, Ni, and other transition metals may limit the waste loading and affect the durability of the glass by forming spine1 phases at the high operating temperature used in vitrification. Some magnetic spine1 phases observed in glass formation are magnetite (Fe,O,), chromite (FeCrO,), and others [(Fe, Ni, Mg, Zn, Mn)(Al, Fe, Ti, Cr)O,] as described elsewhere [Bates-1994, Wronkiewicz-1994] Stable spine1 phases can cause segregation between the glass and the crystalline phases. As a consequence of the difference in density, the spine1 phases tend to accumulate at the bottom of the glass melter, which decreases the conductivity and melter lifetime [Sproull-1993]. Crystallization also can affect glass durability [Jantzen-1985, Turcotte- 1979, Buechele-1990] by changing the chemical composition of the matrix glass surrounding the crystals or causing stress at the glass/crystal interface. These are some of the effects that can increase leaching [Jantzen-1985]. A SRS glass that was partially crystallized to contain 10% vol. crystals composed of spinels, nepheline, and acmite phases showed minimal changes in

  8. Vitrification of reactor wastes

    International Nuclear Information System (INIS)

    Jouan, A.

    1993-01-01

    The vitrification of low and intermediate level wastes from the NPP operation has been studied in the frame of a Franco-Czech agreement. The laboratory experiments concentrated on a search for a suitable borosilicate glass matrix which could incorporate relatively high quantities of boron and sodium, main components of liquid wastes from the WWER reactor type NPPs. A relatively wide area of waste compositions has been studied and properties of glasses suitable for the technology and waste disposal were measured. Great attention has been paid to the chemical stability (leachability), other properties like thermal dependence of viscosity and electrical conductivity of melts, and the microstructure of the final solidification product have also been evaluated. The feasibility of the vitrification process has been proved during pilot plant tests which were accomplished at the French establishment in Marcoule. The results of tests were promising. (authors). 4 tabs., 7 figs

  9. Vitrification of reactor wastes

    Energy Technology Data Exchange (ETDEWEB)

    Jouan, A [CEA Centre d` Etudes de la Vallee du Rhone, 30 - Marcoule (France). Dept. des Procedes de Retraitement; Sussmilch, J [Nuclear Research Institut, Rez (Czech Republic)

    1994-12-31

    The vitrification of low and intermediate level wastes from the NPP operation has been studied in the frame of a Franco-Czech agreement. The laboratory experiments concentrated on a search for a suitable borosilicate glass matrix which could incorporate relatively high quantities of boron and sodium, main components of liquid wastes from the WWER reactor type NPPs. A relatively wide area of waste compositions has been studied and properties of glasses suitable for the technology and waste disposal were measured. Great attention has been paid to the chemical stability (leachability), other properties like thermal dependence of viscosity and electrical conductivity of melts, and the microstructure of the final solidification product have also been evaluated. The feasibility of the vitrification process has been proved during pilot plant tests which were accomplished at the French establishment in Marcoule. The results of tests were promising. (authors). 4 tabs., 7 figs.

  10. Vitrification of surrogate mixed wastes in a graphite electrode arc melter

    International Nuclear Information System (INIS)

    Soelberg, N.R.; Chambers, A.G.; Ball, L.

    1995-01-01

    Demonstration tests for vitrifying mixed wastes and contaminated soils have been conducted using a small (800 kVA), industrial-scale, three-phase AC, graphite electrode furnace located at the Albany Research Center of the United States Bureau of Mines (USBM). The feed mixtures were non-radioactive surrogates of various types of mixed (radioactive and hazardous), transuranic-contaminated wastes stored and buried at the Idaho National Engineering Laboratory (INEL). The feed mixtures were processed with added soil from the INEL. Objectives being evaluated include (1) equipment capability to achieve desired process conditions and vitrification products for different feed compositions, (2) slag and metals tapping capability, (3) partitioning of transuranic elements and toxic metals among the furnace products, (4) slag, fume, and metal products characteristics, and (5) performance of the feed, furnace and air pollution control systems. The tests were successfully completed in mid-April 1995. A very comprehensive process monitoring, sampling and analysis program was included in the test program. Sample analysis, data reduction, and results evaluation are currently underway. Initial results indicate that the furnace readily processed around 20,000 lb of widely ranging feed mixtures at feedrates of up to 1,100 lb/hr. Continuous feeding and slag tapping was achieved. Molten metal was also tapped twice during the test program. Offgas emissions were efficiently controlled as expected by a modified air pollution control system

  11. Vitrification chemistry and nuclear waste

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1985-01-01

    The vitrification of nuclear waste offers unique challenges to the glass technologist. The waste contains 50 or 60 elements, and often varies widely in composition. Most of these elements are seldom encountered in processing commercial glasses. The melter to vitrify the waste must be able to tolerate these variations in composition, while producing a durable glass. This glass must be produced without releasing hazardous radionuclides to the environment during any step of the vitrification process. Construction of a facility to convert the nearly 30 million gallons of high-level nuclear waste at the Savannah River Plant into borosilicate glass began in late 1983. In developing the vitrification process, the Savannah River Laboratory has had to overcome all of these challenges to the glass technologist. Advances in understanding in three areas have been crucial to our success: oxidation-reduction phenomena during glass melting; the reaction between glass and natural wastes; and the causes of foaming during glass melting

  12. Hanford Waste Vitrification Plant Dangerous Waste Permit Application

    International Nuclear Information System (INIS)

    1991-10-01

    The Hanford Facility currently stores mixed waste, resulting from various processing operations, in underground storage tanks. The Hanford Waste Vitrification Plant will be constructed and operated to process the high-activity fraction of mixed waste stored in these underground tanks. The Hanford Waste Vitrification Plant will solidify pretreated tank waste into a glass product that will be packaged for disposal in a national repository. This Vitrification Plant Dangerous Waste Permit Application, Revision 2, consists of both a Part A and a Part B permit application. An explanation of the Part A revisions, including Revision 4 submitted with this application, is provided at the beginning of the Part A section. The Part B consists of 15 chapters addressing the organization and content of the Part B Checklist prepared by the Washington State Department of Ecology (Ecology 1987)

  13. In situ vitrification of a mixed-waste contaminated soil site: The 116-B-6A crib at Hanford

    Energy Technology Data Exchange (ETDEWEB)

    Luey, J.; Koegler, S.S.; Kuhn, W.L.; Lowery, P.S.; Winkelman, R.G.

    1992-09-01

    The first large-scale mixed-waste test of in situ vitrification (ISV) has been completed. The large-scale test was conducted at an actual contaminated soil site, the 116-B-6A crib, on the Department of Energy's Hanford Site. The large-scale test was a demonstration of the ISV technology and not an interim action for the 116-B-6A crib. This demonstration has provided technical data to evaluate the ISV process for its potential in the final disposition of mixed-waste contaminated soil sites at Hanford. Because of the test's successful completion. technical data on the vitrified soil are available on how well the process incorporates transuranics and heavy metals into the waste form. how well the form resists leaching of transuranics and heavy metals. how well the process handles sites with high combustible loadings, and the important site parameters which may affect the achievable process depth. This report describes the 116-B-6A crib site, the objectives of the ISV demonstration, the results in terms of the objectives, and the overall process performance.

  14. Vitrification of hazardous and radioactive wastes

    International Nuclear Information System (INIS)

    Bickford, D.F.; Schumacher, R.

    1995-01-01

    Vitrification offers many attractive waste stabilization options. Versatility of waste compositions, as well as the inherent durability of a glass waste form, have made vitrification the treatment of choice for high-level radioactive wastes. Adapting the technology to other hazardous and radioactive waste streams will provide an environmentally acceptable solution to many of the waste challenges that face the public today. This document reviews various types and technologies involved in vitrification

  15. Glasses and nuclear waste vitrification

    International Nuclear Information System (INIS)

    Ojovan, Michael I.

    2012-01-01

    Glass is an amorphous solid material which behaves like an isotropic crystal. Atomic structure of glass lacks long-range order but possesses short and most probably medium range order. Compared to crystalline materials of the same composition glasses are metastable materials however crystallisation processes are kinetically impeded within times which typically exceed the age of universe. The physical and chemical durability of glasses combined with their high tolerance to compositional changes makes glasses irreplaceable when hazardous waste needs immobilisation for safe long-term storage, transportation and consequent disposal. Immobilisation of radioactive waste in glassy materials using vitrification has been used successfully for several decades. Nuclear waste vitrification is attractive because of its flexibility, the large number of elements which can be incorporated in the glass, its high corrosion durability and the reduced volume of the resulting wasteform. Vitrification involves melting of waste materials with glass-forming additives so that the final vitreous product incorporates the waste contaminants in its macro- and micro-structure. Hazardous waste constituents are immobilised either by direct incorporation into the glass structure or by encapsulation when the final glassy material can be in form of a glass composite material. Both borosilicate and phosphate glasses are currently used to immobilise nuclear wastes. In addition to relatively homogeneous glasses novel glass composite materials are used to immobilise problematic waste streams. (author)

  16. Vitrification of copper flotation waste

    Energy Technology Data Exchange (ETDEWEB)

    Karamanov, Alexander [Institute of Physical Chemistry, Bulgarian Academy of Science, G. Bonchev Str. Block 11, 1113 Sofia (Bulgaria)]. E-mail: karama@ing.univaq.it; Aloisi, Mirko [Department of Chemistry, Chemical Engineering and Materials, University of L' Aquila, 67040 Monteluco di Roio, L' Aquila (Italy); Pelino, Mario [Department of Chemistry, Chemical Engineering and Materials, University of L' Aquila, 67040 Monteluco di Roio, L' Aquila (Italy)

    2007-02-09

    The vitrification of an hazardous iron-rich waste (W), arising from slag flotation of copper production, was studied. Two glasses, containing 30 wt% W were melted for 30 min at 1400 deg. C. The first batch, labeled WSZ, was obtained by mixing W, blast furnace slag (S) and zeolite tuff (Z), whereas the second, labeled WG, was prepared by mixing W, glass cullet (G), sand and limestone. The glass frits showed high chemical durability, measured by the TCLP test. The crystallization of the glasses was evaluated by DTA. The crystal phases formed were identified by XRD resulting to be pyroxene and wollastonite solid solutions, magnetite and hematite. The morphology of the glass-ceramics was observed by optical and scanning electron microscopy. WSZ composition showed a high rate of bulk crystallization and resulted to be suitable for producing glass-ceramics by a short crystallization heat-treatment. WG composition showed a low crystallization rate and good sinterability; glass-ceramics were obtained by sinter-crystallization of the glass frit.

  17. Vitrification of copper flotation waste.

    Science.gov (United States)

    Karamanov, Alexander; Aloisi, Mirko; Pelino, Mario

    2007-02-09

    The vitrification of an hazardous iron-rich waste (W), arising from slag flotation of copper production, was studied. Two glasses, containing 30wt% W were melted for 30min at 1400 degrees C. The first batch, labeled WSZ, was obtained by mixing W, blast furnace slag (S) and zeolite tuff (Z), whereas the second, labeled WG, was prepared by mixing W, glass cullet (G), sand and limestone. The glass frits showed high chemical durability, measured by the TCLP test. The crystallization of the glasses was evaluated by DTA. The crystal phases formed were identified by XRD resulting to be pyroxene and wollastonite solid solutions, magnetite and hematite. The morphology of the glass-ceramics was observed by optical and scanning electron microscopy. WSZ composition showed a high rate of bulk crystallization and resulted to be suitable for producing glass-ceramics by a short crystallization heat-treatment. WG composition showed a low crystallization rate and good sinterability; glass-ceramics were obtained by sinter-crystallization of the glass frit.

  18. Vitrification of copper flotation waste

    International Nuclear Information System (INIS)

    Karamanov, Alexander; Aloisi, Mirko; Pelino, Mario

    2007-01-01

    The vitrification of an hazardous iron-rich waste (W), arising from slag flotation of copper production, was studied. Two glasses, containing 30 wt% W were melted for 30 min at 1400 deg. C. The first batch, labeled WSZ, was obtained by mixing W, blast furnace slag (S) and zeolite tuff (Z), whereas the second, labeled WG, was prepared by mixing W, glass cullet (G), sand and limestone. The glass frits showed high chemical durability, measured by the TCLP test. The crystallization of the glasses was evaluated by DTA. The crystal phases formed were identified by XRD resulting to be pyroxene and wollastonite solid solutions, magnetite and hematite. The morphology of the glass-ceramics was observed by optical and scanning electron microscopy. WSZ composition showed a high rate of bulk crystallization and resulted to be suitable for producing glass-ceramics by a short crystallization heat-treatment. WG composition showed a low crystallization rate and good sinterability; glass-ceramics were obtained by sinter-crystallization of the glass frit

  19. Superconducting open-gradient magnetic separation for the pretreatment of radioactive or mixed waste vitrification feeds. 1997 annual progress report

    International Nuclear Information System (INIS)

    Cicero-Herman, C.A.; Ritter, J.A.

    1997-01-01

    'Vitrification has been selected as a final waste form technology in the US for long-term storage of high-level radioactive wastes (HLW). However, a foreseeable problem during vitrification in some waste feed streams lies in the presence of elements (e.g., transition metals) in the HLW that may cause instabilities in the final glass product. The formation of spinel compounds, such as Fe 3 O 4 and FeCrO 4 , results in glass phase separation and reduces vitrifier lifetime, and durability of the final waste form. A superconducting open gradient magnetic separation (OGMS) system maybe suitable for the removal of the deleterious transition elements (e.g. Fe, Co, and Ni) and other elements (lanthanides) from vitrification feed streams due to their ferromagnetic or paramagnetic nature. The OGMS systems are designed to deflect and collect paramagnetic minerals as they interact with a magnetic field gradient. This system has the potential to reduce the volume of HLW for vitrification and ensure a stable product. In order to design efficient OGMS and High gradient magnetic separation (HGMS) processes, a fundamental understanding of the physical and chemical properties of the waste feed streams is required. Using HLW simulant and radioactive fly ash and sludge samples from the Savannah River Technology Center, Rocky Flats site, and the Hanford reservation, several techniques were used to characterize and predict the separation capability for a superconducting OGMS system.'

  20. Plasma/arc melter review for vitrification of mixed wastes: Results

    Energy Technology Data Exchange (ETDEWEB)

    Eddy, T.L.; Soelberg, N.R.; Raivo, B.D. [MeltTran, Inc., Idaho Falls, ID (United States)

    1995-12-31

    In October of 1994, the Idaho Waste Treatment Program (IWTP) sponsored a workshop to review the results of a plasma/arc melter system preliminary design for treating mixed waste. Attention focused on (1) the melter design, (2) the offgas system design, and (3) the overall system design. The inclusion of feed preparation and handling systems, as well as monitoring and control systems, were considered premature until decisions regarding the melter and offgas treatment were resolved. The evaluation was based on the constraints of the transuranic-contaminated mixed waste in the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering Laboratory (INEL). Major factors are the retention of the transuranics in the basaltic slag, maintenance in a radioactive environment, reliability of components to prevent any major problems, upsets, or safety concerns, and the collection, elimination, or reduction of hazardous materials for appropriate stabilization. Several modifications were recommended by the group at large, discussed by the subcommittees, and accepted as the preferred options by the design team. Though all questions were not answered, the preferred systems for mixed waste treatment were the arc melters with graphite electrode systems with appropriate cooling which reduced maintenance and the possibility of eruptions that have occurred with plasma torches. Arc melters can also result in the minimum footprint and shielding. The preferred offgas systems were the wet/dry systems, that essentially eliminate the formation of carcinogenic compounds so they do not have to be destroyed down stream. This system also puts all of the particulate matter into one stream, instead of two.

  1. Novel procurement concepts utilized to award contract for vitrification of an F006 mixed waste sludge

    International Nuclear Information System (INIS)

    Pickett, J.B.; Musall, J.C.; Hayes, A.F.; Campbell, E.E.

    1994-01-01

    A number of novel concepts were utilized in a procurement bid process to award a contract for the stabilization of a mixed wastewater treatment plating line sludge from the Reactor Materials department (M-Area) at the Savannah River site (SRS). The contract award was based on a combination of technical and cost considerations. The technical aspects included an evaluation of the technical validity of the proposed process(es) (i.e., would the process work?), the physical resources of the proposer and the expertise of the personnel proposed to work on the program, and prior experience of the firm wit treatment and stabilization of mixed (radioactive and hazardous) wastes. This paper will concentrate on the cost and the bid award considerations, rather than the technical aspects

  2. Vitrification of organics-containing wastes

    International Nuclear Information System (INIS)

    Bickford, D.F.

    1997-01-01

    A process is described for stabilizing organics-containing waste materials and recovering metals therefrom, and a waste glass product made according to the process is also disclosed. Vitrification of wastes such as organic ion exchange resins, electronic components and the like can be accomplished by mixing at least one transition metal oxide with the wastes, and, if needed, glass formers to compensate for a shortage of silicates or other glass formers in the wastes. The transition metal oxide increases the rate of oxidation of organic materials in the wastes to improve the composition of the glass-forming mixture: at low temperatures, the oxide catalyzes oxidation of a portion of the organics in the waste; at higher temperatures, the oxide dissolves and the resulting oxygen ions oxidize more of the organics; and at vitrification temperatures, the metal ions conduct oxygen into the melt to oxidize the remaining organics. In addition, the transition metal oxide buffers the redox potential of the glass melt so that metals such as Au, Pt, Ag, and Cu separate from the melt in the metallic state and can be recovered. After the metals are recovered, the remainder of the melt is allowed to cool and may subsequently be disposed of. The product has good leaching resistance and can be disposed of in an ordinary landfill, or, alternatively, used as a filler in materials such as concrete, asphalt, brick and tile. 1 fig

  3. Waste Vitrification Projects Throughout the US Initiated by SRS

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Whitehouse, J.C.; Smith, M.E.; Pickett, J.B.; Peeler, D.K.

    1998-05-01

    Technologies are being developed by the U. S. Department of Energy's (DOE) Nuclear Facility sites to convert high-level, low-level, and mixed wastes to a solid stabilized waste form for permanent disposal. Vitrification is one of the most important and environmentally safest technologies being developed. The Environmental Protection Agency (EPA) has declared vitrification the best demonstrated available technology for high-level radioactive waste and produced a Handbook of Vitrification Technologies for Treatment of Hazardous and Radioactive Waste. The Defense Waste Processing Facility being tested at will soon start vitrifying the high-level waste at. The DOE Office of Technology Development has taken the position that mixed waste needs to be stabilized to the highest level reasonably possible to ensure that the resulting waste forms will meet both current and future regulatory specifications. Vitrification produces durable waste forms at volume reductions up to 97%. Large reductions in volume minimize long-term storage costs making vitrification cost effective on a life cycle basis

  4. In situ vitrification on buried waste

    International Nuclear Information System (INIS)

    Bates, S.O.

    1992-01-01

    In situ vitrification (ISV) is being evaluated as a remedial treatment technology for buried mixed and transuranic (TRU) wastes at the Subsurface Disposal Area (SDA) at Idaho National Engineering Laboratory (INEL) and can be related to buried wastes at other Department of Energy (DOE) sites. There are numerous locations around the DOE Complex where wastes were buried in the ground or stored for future burial. The Buried Waste Program (BWP) is conducting a comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) remedial investigation/feasibility study (RI/FS) for the Department of Energy - Field Office Idaho (DOE-ID). As part of the RI/FS, an ISV scoping study on the treatability of the SDA mixed low-level and mixed TRU waste is being performed for applicability to remediation of the waste at the Radioactive Waste Management Complex (RWMC). The ISV project being conducted at the INEL by EG ampersand G Idaho, Inc. consists of a treatability investigation to collect data to satisfy nine CERCLA criteria with regards to the SDA. This treatability investigation involves a series of experiments and related efforts to study the feasibility of ISV for remediation of mixed and TRU waste disposed of at the SDA

  5. Radioactive waste vitrification: A review

    International Nuclear Information System (INIS)

    Cole, L.L.; Fields, D.E.

    1989-08-01

    The research and development of an immobilization process for the containment of nuclear high-level liquid waste has been underway for well-over the past four decades. The method that has become the state-of-the-art is the liquid-fed ceramic melter process which converts a mixture of high-level liquid waste and glass forming frit to a borosilicate glass product. This report gives a chronological review of the various vitrification processes starting with the very first reported process in 1960. Information on the early methods of frit selection as well as information on the currently computerized method are presented. The importance of all these parameters is discussed with regard to product durability. 26 refs., 8 figs., 1 tab

  6. Superconducting open-gradient magnetic separation for the pretreatment of radioactive or mixed waste vitrification feeds. 1998 annual progress report

    International Nuclear Information System (INIS)

    Crawford, C.; Doctor, R.D.; Landsberger, S.; Nunez, L.; Ritter, J.

    1998-01-01

    'The objective is to reduce the volume and cost of high-level waste glass produced during US DOE remediation activities by demonstrating that magnetic separation can separate crystalline, amorphous, and colloidal constituents in vitrification feed streams known to be deleterious to the production of borosilicate glass. Magnetic separation will add neither chemicals nor generate secondary waste streams. The project includes the systematic study of magnetic interactions of waste constituents under controlled physical and chemical conditions (e.g., hydration, oxidation, temperature) to identify mechanisms that control the magnetic properties. Partitioning of radionuclides to determine their sorption mechanisms is also being studied. The identification of fundamental magnetic properties within the microscopic chemical environment in combination with hydrodynamic and electrodynamic models provides insights into the design of a system for optimal separation. Following this, experimental studies using superconducting open-gradient magnetic separation (OGMS) will be conducted to validate its effectiveness as a pretreatment technique.'

  7. High-Level Waste Vitrification Facility Feasibility Study

    International Nuclear Information System (INIS)

    D. A. Lopez

    1999-01-01

    A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035

  8. High-Level Waste Vitrification Facility Feasibility Study

    Energy Technology Data Exchange (ETDEWEB)

    D. A. Lopez

    1999-08-01

    A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035.

  9. Hanford waste vitrification systems risk assessment

    International Nuclear Information System (INIS)

    Miller, W.C.; Hamilton, D.W.; Holton, L.K.; Bailey, J.W.

    1991-09-01

    A systematic Risk Assessment was performed to identify the technical, regulatory, and programmatic uncertainties and to quantify the risks to the Hanford Site double-shell tank waste vitrification program baseline (as defined in December 1990). Mitigating strategies to reduce the overall program risk were proposed. All major program elements were evaluated, including double-shell tank waste characterization, Tank Farms, retrieval, pretreatment, vitrification, and grouting. Computer-based techniques were used to quantify risks to proceeding with construction of the Hanford Waste Vitrification Plant on the present baseline schedule. Risks to the potential vitrification of single-shell tank wastes and cesium and strontium capsules were also assessed. 62 refs., 38 figs., 26 tabs

  10. Waste vitrification: a historical perspective

    International Nuclear Information System (INIS)

    McElroy, J.L.; Bjorklund, W.J.; Bonner, W.F.

    1982-02-01

    The possibility of converting high-level wastes (HLW) to glass was first pursued in Canada and England at a time when other countries were evaluating many other alternatives. By 1966, the British had completed radioactive demonstrations of the FINGAL pot process, converting HLW to borosilicate glass. By this time other countries, including France and the United States, had begun using the glass waste form. Beginning in 1966, several processes, including phosphate and borosilicate glass, were demonstrated by the US in the Waste Solidification Engineering Prototypes (WSEP) program at the Pacific Northwest Laboratory (PNL). Most of the current vitrification processes are adaptations of the FINGAL pot process or the continuous metallic melter used in the WSEP program. One notable exception is the joule-heated ceramic melter, which was adapted from commercial glass technology for HLW by PNL in the mid-1970's. Both batch and continuous processes have been developed to an advanced stage of readiness. These processes are described and compared in this paper

  11. Hanford Waste Vitrification Plant Quality Assurance Program description for high-level waste form development and qualification

    International Nuclear Information System (INIS)

    1993-08-01

    The Hanford Waste Vitrification Plant Project has been established to convert the high-level radioactive waste associated with nuclear defense production at the Hanford Site into a waste form suitable for disposal in a deep geologic repository. The Hanford Waste Vitrification Plant will mix processed radioactive waste with borosilicate material, then heat the mixture to its melting point (vitrification) to forin a glass-like substance that traps the radionuclides in the glass matrix upon cooling. The Hanford Waste Vitrification Plant Quality Assurance Program has been established to support the mission of the Hanford Waste Vitrification Plant. This Quality Assurance Program Description has been written to document the Hanford Waste Vitrification Plant Quality Assurance Program

  12. Treatment of NPP wastes using vitrification

    International Nuclear Information System (INIS)

    Sobolev, I.A.; Lifanov, F.A.; Stefanovsky, S.V.; Kobelev, A.P.; Savkin, A.E.; Kornev, V.I.

    1998-01-01

    Glass-based materials to immobilize various liquid and solid radioactive wastes generated at nuclear power plants (NPP) were designed. Glassy waste forms can be produced using electric melting including a cold crucible melting. Leach rate of cesium was found to be 10 -5 -10 -6 g/(cm 2 day) (IAEA technique). Volume reduction factor after vitrification reached 4-5. Various technologies for NPP waste vitrification were developed. Direct vitrification means feeding of source waste into the melter with formation of glassy waste form to be disposed. Joule heated ceramic melter, and cold crucible were tested. Process variables at treatment of Kursk, Chernobyl (RBMK), Kalinin, Novovoronezh (VVER) NPP wastes were determined. The most promising melter was found to be the cold crucible. Pilot plant based on the cold crucibles has been designed and constructed. Solid burnable NPP wastes are incinerated and slags are incorporated in glass. (author)

  13. Vitrification process testing for reference HWVP waste

    International Nuclear Information System (INIS)

    Perez, J.M. Jr.; Goles, R.W.; Nakaoka, R.K.; Kruger, O.L.

    1991-01-01

    The Hanford Waste Vitrification Plant (HWVP) is being designed to vitrify high-level radioactive wastes stored on the Hanford site. The vitrification flow-sheet is being developed to assure the plant will achieve plant production requirements and the glass product will meet all waste form requirements for final geologic disposal. The first Hanford waste to be processed by the HWVP will be a neutralized waste resulting from PUREX fuel reprocessing operations. Testing is being conducted using representative nonradioactive simulants to obtain process and product data required to support design, environmental, and qualification activities. Plant/process criteria, testing requirements and approach, and results to date will be presented

  14. In situ vitrification: Application to buried waste

    International Nuclear Information System (INIS)

    Callow, R.A.; Thompson, L.E.

    1991-01-01

    Two in situ vitrification field tests were conducted in June and July 1990 at Idaho National Engineering Laboratory. In situ vitrification is a technology for in-place conversion of contaminated soils into a durable glass and crystalline waste form and is being investigated as a potential remediation technology for buried waste. The overall objective of the two tests was to assess the general suitability of the process to remediate buried waste structures found at Idaho National Engineering Laboratory. In particular, these tests were designed as part of a treatability study to provide essential information on 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 provided valuable operational control for successfully processing the high metal content waste. The results indicate that in situ vitrification is a feasible technology for application to buried waste. 2 refs., 5 figs., 2 tabs

  15. Method of vitrificating fine-containing liquid waste

    International Nuclear Information System (INIS)

    Hagiwara, Minoru; Matsunaka, Kazuhisa.

    1989-01-01

    This invention concerns a vitrificating method of liquid wastes containing fines (metal powder discharged upon cutting fuel cans) used in a process for treating high level radioactive liquid wastes or a process for treating liquid wastes from nuclear power plants. Liquid wastes containing fines, slurries, etc. are filtered by a filter vessel comprising glass fibers. The fines are supplied as they are to a glass melting furnace placed in the vessel. Filterates formed upon filteration are mixed with other high level radioactive wastes and supplied together with starting glass material to the glass melting furnace. Since the fine-containing liquid wastes are processed separately from high radioactive liquid wastes, clogging of pipeways, etc. can be avoided, supply to the melting furnace is facilitated and the operation efficiency of the vitrification process can be improved. (I.N.)

  16. Plasma vitrification of waste materials

    Science.gov (United States)

    McLaughlin, David F.; Dighe, Shyam V.; Gass, William R.

    1997-01-01

    This invention provides a process wherein hazardous or radioactive wastes in the form of liquids, slurries, or finely divided solids are mixed with finely divided glassformers (silica, alumina, soda, etc.) and injected directly into the plume of a non-transferred arc plasma torch. The extremely high temperatures and heat transfer rates makes it possible to convert the waste-glassformer mixture into a fully vitrified molten glass product in a matter of milliseconds. The molten product may then be collected in a crucible for casting into final wasteform geometry, quenching in water, or further holding time to improve homogeneity and eliminate bubbles.

  17. Plasma vitrification of waste materials

    International Nuclear Information System (INIS)

    McLaughlin, D.F.; Dighe, S.V.; Gass, W.R.

    1997-01-01

    This invention provides a process wherein hazardous or radioactive wastes in the form of liquids, slurries, or finely divided solids are mixed with finely divided glassformers (silica, alumina, soda, etc.) and injected directly into the plume of a non-transferred arc plasma torch. The extremely high temperatures and heat transfer rates makes it possible to convert the waste-glassformer mixture into a fully vitrified molten glass product in a matter of milliseconds. The molten product may then be collected in a crucible for casting into final wasteform geometry, quenching in water, or further holding time to improve homogeneity and eliminate bubbles. 4 figs

  18. Hanford Waste Vitrification Plant technology progress

    International Nuclear Information System (INIS)

    Wolfe, B.A.; Scott, J.L.; Allen, C.R.

    1989-10-01

    The Hanford Waste Vitrification Plant (HWVP) is currently being designed to safely process and temporarily store immobilized defense liquid high-level wastes from the Hanford Site. These wastes will be immobilized in a borosilicate glass waste form in the HWVP and stored onsite until a qualified geologic waste repository is ready for permanent disposal. Because of the diversity of wastes to be disposed of, specific technical issues are being addressed so that the plant can be designed and operated to produce a waste form that meets the requirements for permanent disposal in a geologic repository. This paper reports the progress to date in addressing these issues. 2 figs., 3 tabs

  19. Vitrification technology for Hanford Site tank waste

    International Nuclear Information System (INIS)

    Weber, E.T.; Calmus, R.B.; Wilson, C.N.

    1995-04-01

    The US Department of Energy's (DOE) Hanford Site has an inventory of 217,000 m 3 of nuclear waste stored in 177 underground tanks. The DOE, the US Environmental Protection Agency, and the Washington State Department of Ecology have agreed that most of the Hanford Site tank waste will be immobilized by vitrification before final disposal. This will be accomplished by separating the tank waste into high- and low-level fractions. Capabilities for high-capacity vitrification are being assessed and developed for each waste fraction. This paper provides an overview of the program for selecting preferred high-level waste melter and feed processing technologies for use in Hanford Site tank waste processing

  20. Hanford Waste Vitrification Plant Technology Plan

    International Nuclear Information System (INIS)

    Sexton, R.A.

    1988-06-01

    The reference Hanford plan for disposal of defense high-level waste is based on waste immobilization in glass by the vitrification process and temporary vitrified waste storage at the Hanford Site until final disposal in a geologic repository. A companion document to the Hanford Waste Management Plan (HWMP) is the Draft, Interim Hanford Waste Management Technology Plan (HWMTP), which provides a description of the technology that must be developed to meet the reference waste management plan. One of the issues in the HWMTP is DST-6, Immobilization (Glass). The HWMTP includes all expense funding needed to complete the Hanford Waste Vitrification Plant (HWVP) project. A preliminary HWVP Technology Plan was prepared in 1985 as a supporting document to the HWMTP to provide a more detailed description of the technology needed to construct and operate a vitrification facility. The plan was updated and issued in 1986, and revised in 1987. This document is an annual update of the plan. The HWVP Technology Plan is limited in scope to technology that requires development or confirmation testing. Other expense-funded activities are not included. The relationship between the HWVP Technology Plan and other waste management issues addressed in the HWMTP is described in section 1.6 of this plan. 6 refs., 4 figs., 34 tabs

  1. Hanford Waste Vitrification Plant applied technology plan

    International Nuclear Information System (INIS)

    Kruger, O.L.

    1990-09-01

    This Applied Technology Plan describes the process development, verification testing, equipment adaptation, and waste form qualification technical issues and plans for resolution to support the design, permitting, and operation of the Hanford Waste Vitrification Plant. The scope of this Plan includes work to be performed by the research and development contractor, Pacific Northwest Laboratory, other organizations within Westinghouse Hanford Company, universities and companies with glass technology expertise, and other US Department of Energy sites. All work described in this Plan is funded by the Hanford Waste Vitrification Plant Project and the relationship of this Plan to other waste management documents and issues is provided for background information. Work to performed under this Plan is divided into major areas that establish a reference process, develop an acceptable glass composition envelope, and demonstrate feed processing and glass production for the range of Hanford Waste Vitrification Plant feeds. Included in this work is the evaluation and verification testing of equipment and technology obtained from the Defense Waste Processing Facility, the West Valley Demonstration Project, foreign countries, and the Hanford Site. Development and verification of product and process models and other data needed for waste form qualification documentation are also included in this Plan. 21 refs., 4 figs., 33 tabs

  2. Hanford Waste Vitrification Project overview and status

    International Nuclear Information System (INIS)

    Swenson, L.D.; Smets, J.L.

    1993-01-01

    The Hanford Waste Vitrification Project (HWVP) is being constructed at the US DOE's Hanford Site in Richland, WA. Engineering and design are being accomplished by Fluor Daniel Inc. in Irvine, CA. Technical input is furnished by Westinghouse Hanford Co. and construction management services by UE ampersand C-Catalytic Inc. The HWVP will immobilize high level nuclear waste in a glass matrix for eventual disposal in the federal repository. The HWVP consists of several structures, the major ones being the Vitrification Building, the Canister Storage Building, fan house, sand filter, waste hold tank, pump house, and administration and construction facilities. Construction started in April 1992 with the clearing and grubbing activities that prepared the site for fencing and construction preparation. Several design packages have been released for procurement activities. The most significant package release is for the Canister Storage Building, which will be the first major structure to be constructed

  3. Vitrification of high level wastes in France

    International Nuclear Information System (INIS)

    Sombret, C.

    1984-02-01

    A brief historical background of the research and development work conducted in France over 25 years is first presented. Then, the papers deals with the vitrification at (1) the UP1 reprocessing plant (Marcoule) and (2) the UP2 and UP3 reprocessing plants (La Hague). 1) The properties of glass required for high-level radioactive waste vitrification are recalled. The vitrification process and facility of Marcoule are presented. (2) The average characteristics (chemical composition, activity) of LWR fission product solution are given. The glass formulations developed to solidify LWR waste solution must meet the same requirements as those used in the UP1 facility at Marcoule. Three important aspects must be considered with respect to the glass fabrication process: corrosiveness of the molten glass with regard to metals, viscosity of the molten glass, and, volatization during glass fabrication. The glass properties required in view of interim storage and long-term disposal are then largely developed. Two identical vitrification facilities are planned for the site: T7, to process the UP2 throughput, and T7 for the UP3 plant. A prototype unit was built and operated at Marcoule

  4. Hanford Waste Vitrification Plant capacity increase options

    International Nuclear Information System (INIS)

    Larson, D.E.

    1996-04-01

    Studies are being conducted by the Hanford Waste Vitrification Plant (HWVP) Project on ways to increase the waste processing capacity within the current Vitrification Building structural design. The Phase 1 study on remote systems concepts identification and extent of capacity increase was completed. The study concluded that the HWVP capacity could be increased to four times the current capacity with minor design adjustments to the fixed facility design, and the required design changes would not impact the current footprint of the vitrification building. A further increase in production capacity may be achievable but would require some technology development, verification testing, and a more systematic and extensive engineering evaluation. The primary changes included a single advance melter with a higher capacity, new evaporative feed tank, offgas quench collection tank, ejector venturi scrubbers, and additional inner canister closure station,a smear test station, a new close- coupled analytical facility, waste hold capacity of 400,000 gallon, the ability to concentrate out-of-plant HWVP feed to 90 g/L waste oxide concentration, and limited changes to the current base slab construction package

  5. Plasma vitrification program for radioactive waste treatment

    International Nuclear Information System (INIS)

    Hung, Tsungmin; Tzeng, Chinchin; Kuo, Pingchun

    1998-01-01

    In order to treat radioactive wastes effectively and solve storage problems, INER has developed the plasma arc technology and plasma process for various waste forms for several years. The plasma vitrification program is commenced via different developing stages through nine years. It includes (a) development of non-transferred DC plasma torch, (b) establishment of a lab-scale plasma system with home-made 100kW non-transferred DC plasma torch, (c) testing of plasma vitrification of simulated radioactive wastes, (d) establishment of a transferred DC plasma torch delivering output power more than 800 kW, (e) study of NOx reduction process for the plasma furnace, (f) development of a pilot-scale plasma melting furnace to verify the vitrification process, and (g) constructing a plasma furnace facility in INER. The final goal of the program is to establish a plasma processing plant with capacity of 250 kg/hr to treat the low-level radioactive wastes generated from INER itself and domestic institutes due to isotope applications. (author)

  6. Vitrification: a solution for the wastes of wastes; La vitrification: ca chauffe pour les ultimes

    Energy Technology Data Exchange (ETDEWEB)

    Guihard, B. [Europlasma, 33 - Saint Medard en Jalles (France)

    1997-07-01

    The incineration of wastes generates other wastes (fly ashes) that concentrate a large amount of polluting substances (heavy metals, salts..). French law requires a stabilization of this kind of wastes before their storage. Today vitrification can be considered as an alternative to the stabilization and storage way, the vitrified products could be seen as an interesting material in the building industry or in road works. A few years ago the municipality of Bordeaux decided to launch a demonstration program and a REFIOM (fly ashes) vitrification unit has been operating since 1997. (A.C.)

  7. In situ vitrification of buried waste sites

    International Nuclear Information System (INIS)

    Shade, J.W.; Thompson, L.E.; Kindle, C.H.

    1991-04-01

    In situ vitrification (ISV) is a remedial technology initially developed to treat soils contaminated with a variety of organics, heavy metals, and/or radioactive materials. Recent tests have indicated the feasibility of applying the process to buried wastes including containers, combustibles, and buried metals. In addition, ISV is being considered for application to the emplacement of barriers and to the vitrification of underground tanks. This report provides a review of some of the recent experiences of applying ISV in engineering-scale and pilot-scale tests to wastes containing organics, the Environmental Protection Agency (EPA) Toxic metals buried in sealed containers, and buried ferrous metals, with emphasis on the characteristics of the vitrified product and adjacent soil. 9 refs., 2 figs., 3 tabs

  8. Hanford Waste Vitrification Plant Quality Assurance Program description for high-level waste form development and qualification. Revision 3, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    The Hanford Waste Vitrification Plant Project has been established to convert the high-level radioactive waste associated with nuclear defense production at the Hanford Site into a waste form suitable for disposal in a deep geologic repository. The Hanford Waste Vitrification Plant will mix processed radioactive waste with borosilicate material, then heat the mixture to its melting point (vitrification) to forin a glass-like substance that traps the radionuclides in the glass matrix upon cooling. The Hanford Waste Vitrification Plant Quality Assurance Program has been established to support the mission of the Hanford Waste Vitrification Plant. This Quality Assurance Program Description has been written to document the Hanford Waste Vitrification Plant Quality Assurance Program.

  9. Chemical engineering problems of radioactive waste fixation by vitrification

    International Nuclear Information System (INIS)

    Taylor, R.F.

    1985-01-01

    Basic features are reviewed of the chemical engineering problems faced in the vitrification of the high-level radioactive liquid wastes resulting from the reprocessing of nuclear fuel. After an outline of glass solution properties and formation kinetics the constituent elements of the vitrification route are examined in turn: waste feed evaporation and denitration, calcination, offgas treatment, and finally melting and product quality. Plant and experimental data for each stage are discussed with comparison between process routes and with reference to the underlying principles. Attention is drawn to the future need for higher trapping efficiencies and for dealing with a wider range of species in offgas treatments as higher burnup fuels are processed after shorter cooling times from reactor. Two areas of present study where deeper insight into underlying process mechanics is needed are, firstly, the association of waste material with glass formers in the wet or sinter stages and secondly their incorporation and mixing reaction in the melt. Fuller understanding here would bring direct benefit to process performance and handling. The problems discussed are not of a nature to jeopardize the vitrification routes but if product quality does come to rely heavily on process control then demonstrable confidence in the behaviour of the central physico-chemical interactions is indispensable. (author)

  10. Improved mixing and sampling systems for vitrification melter feeds

    International Nuclear Information System (INIS)

    Ebadian, M.A.

    1998-01-01

    This report summarizes the methods used and results obtained during the progress of the study of waste slurry mixing and sampling systems during fiscal year 1977 (FY97) at the Hemispheric Center for Environmental Technology (HCET) at Florida International University (FIU). The objective of this work is to determine optimal mixing configurations and operating conditions as well as improved sampling technology for defense waste processing facility (DWPF) waste melter feeds at US Department of Energy (DOE) sites. Most of the research on this project was performed experimentally by using a tank mixing configuration with different rotating impellers. The slurry simulants for the experiments were prepared in-house based on the properties of the DOE sites' typical waste slurries. A sampling system was designed to withdraw slurry from the mixing tank. To obtain insight into the waste mixing process, the slurry flow in the mixing tank was also simulated numerically by applying computational fluid dynamics (CFD) methods. The major parameters investigated in both the experimental and numerical studies included power consumption of mixer, mixing time to reach slurry uniformity, slurry type, solids concentration, impeller type, impeller size, impeller rotating speed, sampling tube size, and sampling velocities. Application of the results to the DWPF melter feed preparation process will enhance and modify the technical base for designing slurry transportation equipment and pipeline systems. These results will also serve as an important reference for improving waste slurry mixing performance and melter operating conditions. These factors will contribute to an increase in the capability of the vitrification process and the quality of the waste glass

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

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

  12. Mixed waste: Proceedings

    International Nuclear Information System (INIS)

    Moghissi, A.A.; Blauvelt, R.K.; Benda, G.A.; Rothermich, N.E.

    1993-01-01

    This volume contains the peer-reviewed and edited versions of papers submitted for presentation a the Second International Mixed Waste Symposium. Following the tradition of the First International Mixed Waste Symposium, these proceedings were prepared in advance of the meeting for distribution to participants. The symposium was organized by the Mixed Waste Committee of the American Society of Mechanical Engineers. The topics discussed at the symposium include: stabilization technologies, alternative treatment technologies, regulatory issues, vitrification technologies, characterization of wastes, thermal technologies, laboratory and analytical issues, waste storage and disposal, organic treatment technologies, waste minimization, packaging and transportation, treatment of mercury contaminated wastes and bioprocessing, and environmental restoration. Individual abstracts are catalogued separately for the data base

  13. Mixed waste: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Moghissi, A.A.; Blauvelt, R.K.; Benda, G.A.; Rothermich, N.E. [eds.] [Temple Univ., Philadelphia, PA (United States). Dept. of Environmental Safety and Health

    1993-12-31

    This volume contains the peer-reviewed and edited versions of papers submitted for presentation a the Second International Mixed Waste Symposium. Following the tradition of the First International Mixed Waste Symposium, these proceedings were prepared in advance of the meeting for distribution to participants. The symposium was organized by the Mixed Waste Committee of the American Society of Mechanical Engineers. The topics discussed at the symposium include: stabilization technologies, alternative treatment technologies, regulatory issues, vitrification technologies, characterization of wastes, thermal technologies, laboratory and analytical issues, waste storage and disposal, organic treatment technologies, waste minimization, packaging and transportation, treatment of mercury contaminated wastes and bioprocessing, and environmental restoration. Individual abstracts are catalogued separately for the data base.

  14. In situ vitrification applications to hazardous wastes

    International Nuclear Information System (INIS)

    Liikala, S.

    1989-01-01

    In Situ Vitrification is a new hazardous waste remediation alternative that should be considered for contaminated soil matrices. According to the authors the advantages of using ISV include: technology demonstrated at field scale; applicable to a wide variety of soils and contaminants; pyrolyzer organics and encapsulates inorganics; product durable over geologic time period; no threat of harm to the public from exposure; and applications available for barrier walls and structural support. The use of ISV on a large scale basis has thus far been limited to the nuclear industry but has tremendous potential for widespread applications to the hazardous waste field. With the ever changing regulations for the disposal of hazardous waste in landfills, and the increasing positive analytical data of ISV, the process will become a powerful source for on-site treatment and hazardous waste management needs in the very near future

  15. Preliminary Hanford Waste Vitrification Plan Waste Form Qualification Plan

    International Nuclear Information System (INIS)

    Nelson, J.L.

    1987-09-01

    This Waste Form Qualification Plan describes the waste form qualification activities that will be followed during the design and operation of the Hanford Waste Vitrification Plant to ensure that the vitrified Hanford defense high-level wastes will meet the acceptance requirements of the candidate geologic repositories for nuclear waste. This plan is based on the defense waste processing facility requirements. The content of this plan is based on the assumption that the Hanford Waste Vitrification Plant high-level waste form will be disposed of in one of the geologic repository projects. Proposed legislation currently under consideration by Congress may change or delay the repository site selection process. The impacts of this change will be assessed as details of the new legislation become available. The Plan describes activities, schedules, and programmatic interfaces. The Waste Form Qualification Plan is updated regularly to incorporate Hanford Waste Vitrification Plant-specific waste acceptance requirements and to serve as a controlled baseline plan from which changes in related programs can be incorporated. 10 refs., 5 figs., 5 tabs

  16. Engineering report of plasma vitrification of Hanford tank wastes

    International Nuclear Information System (INIS)

    Hendrickson, D.W.

    1995-01-01

    This document provides an analysis of vendor-derived testing and technology applicability to full scale glass production from Hanford tank wastes using plasma vitrification. The subject vendor testing and concept was applied in support of the Hanford LLW Vitrification Program, Tank Waste Remediation System

  17. Hanford Waste Vitrification Plant hydrogen generation

    International Nuclear Information System (INIS)

    King, R.B.; King, A.D. Jr.; Bhattacharyya, N.K.

    1996-02-01

    The most promising method for the disposal of highly radioactive nuclear wastes is a vitrification process in which the wastes are incorporated into borosilicate glass logs, the logs are sealed into welded stainless steel canisters, and the canisters are buried in suitably protected burial sites for disposal. The purpose of the research supported by the Hanford Waste Vitrification Plant (HWVP) project of the Department of Energy through Battelle Pacific Northwest Laboratory (PNL) and summarized in this report was to gain a basic understanding of the hydrogen generation process and to predict the rate and amount of hydrogen generation during the treatment of HWVP feed simulants with formic acid. The objectives of the study were to determine the key feed components and process variables which enhance or inhibit the.production of hydrogen. Information on the kinetics and stoichiometry of relevant formic acid reactions were sought to provide a basis for viable mechanistic proposals. The chemical reactions were characterized through the production and consumption of the key gaseous products such as H 2 . CO 2 , N 2 0, NO, and NH 3 . For this mason this research program relied heavily on analyses of the gases produced and consumed during reactions of the HWVP feed simulants with formic acid under various conditions. Such analyses, used gas chromatographic equipment and expertise at the University of Georgia for the separation and determination of H 2 , CO, CO 2 , N 2 , N 2 O and NO

  18. Hanford Waste Vitrification Plant Project Waste Form Qualification Program Plan

    International Nuclear Information System (INIS)

    Randklev, E.H.

    1993-06-01

    The US Department of Energy has created a waste acceptance process to help guide the overall program for the disposal of high-level nuclear waste in a federal repository. This Waste Form Qualification Program Plan describes the hierarchy of strategies used by the Hanford Waste Vitrification Plant Project to satisfy the waste form qualification obligations of that waste acceptance process. A description of the functional relationship of the participants contributing to completing this objective is provided. The major activities, products, providers, and associated scheduling for implementing the strategies also are presented

  19. Buried waste remediation: A new application for in situ vitrification

    International Nuclear Information System (INIS)

    Kindle, C.H.; Thompson, L.E.

    1991-04-01

    Buried wastes represent a significant environmental concern and a major financial and technological challenge facing many private firms, local and state governments, and federal agencies. Numerous radioactive and hazardous mixed buried waste sites managed by the US Department of Energy (DOE) require timely clean up to comply with state or federal environmental regulations. Hazardous wastes, biomedical wastes, and common household wastes disposed at many municipal landfills represent a significant environmental health concern. New programs and regulations that result in a greater reduction of waste via recycling and stricter controls regarding generation and disposal of many wastes will help to stem the environmental consequences of wastes currently being generated. Groundwater contamination, methane generation, and potential exposures to biohazards and chemically hazardous materials from inadvertent intrusion will continue to be potential environmental health consequences until effective and permanent closure is achieved. In situ vitrification (ISV) is being considered by the DOE as a permanent closure option for radioactive buried waste sites. The results of several ISV tests on simulated and actual buried wastes conducted during 1990 are presented here. The test results illustrate the feasibility of the ISV process for permanent remediation and closure of buried waste sites in commercial landfills. The tests were successful in immobilizing or destroying hazardous and radioactive contaminants while providing up to 75 vol % waste reduction. 6 refs., 7 figs., 5 tabs

  20. Plasma arc and cold crucible furnace vitrification for medium level waste: a review

    International Nuclear Information System (INIS)

    Poitou, S.; Fiquet, O.; Bourdeloie, C.; Gramondi, P.; Rebollo, F.; Girold, C.; Charvillat, J.P.; Boen, R.; Jouan, A.; Ladirat, C.; Nabot, J.P.; Ochem, D.; Baronnet, J.M.

    2001-01-01

    Initially developed for high-level waste reprocessing, several vitrification processes have been under study since the 80's at the French Atomic Energy Commission (CEA) for other waste categories. According to the French law concerning waste management research passed on December 30, 1991, vitrification may be applied to mixed medium-level waste. A review of processes developed at CEA is presented: cold crucible furnace heated by induced current, refractory furnace heated by nitrogen transferred arc plasma torch, and coupling of cold crucible furnace with oxygen transferred plasma arc twin torch. Furthermore, gas post-combustion has been studied with an oxygen non-transferred plasma torch. (authors)

  1. Hanford Waste Vitrification Plant Clean Air Act permit application

    International Nuclear Information System (INIS)

    1990-04-01

    This document briefly describes the Hanford Site and provides a general overview of the Hanford Waste Vitrification Plant (HWVP). Other topics include sources of emissions, facility operating parameters, facility emissions, pollutant and radionuclide control technology and air quality. The HWVP will convert mixed wastes (high-activity radioactive and hazardous liquid wastes) to a solid vitrified form (borosilicate glass) for disposal. Mixed wastes pretreated in the Hanford Site B Plant will be pumped into double- shell tanks in the 200 East Area for interim storage. This pretreated mixed waste will be batch transferred from interim storage to the HWVP facility, where the waste will be concentrated by evaporation, treated with chemicals, and mixed with glass-forming materials. The mixture will then be continuously fed into an electrically heated glass melter. The molten glass will be poured into canisters that will be cooled, sealed, decontaminated, and stored until the vitrified product can be transferred to a geologic repository. 25 refs., 18 figs., 32 tabs

  2. Vitrification of radioactive waste. Application to other kinds of waste

    International Nuclear Information System (INIS)

    Jouan, A.

    1993-01-01

    The containment by vitrification of radioactive waste is applied to concentrate solutions of fission products coming from the spent fuel reprocessing. By the way of liquid state to solid state, it is possible to reduce the volume of waste, to get a material with safety guarantees necessary to long storage and the glass by its chemical resistance, its thermal stability and its well resistance to irradiation answers particularly well to these necessities

  3. Method for ultimate disposition of borate containing radioactive wastes by vitrification

    International Nuclear Information System (INIS)

    Bege, D.; Faust, H.J.; Puthawala, A.; Stunkel, H.

    1984-01-01

    Method for the ultimate disposition of radioactive wastes by vitrification, in which weak to medium radioactive waste concentrates from borate-containing radioactive liquids are mixed with added glass-forming materials, maximally in a ratio of 1:3, and the mixture heated to obtain a glass-forming melt

  4. Waste vitrification: prediction of acceptable compositions in a lime-soda-silica glass-forming system

    International Nuclear Information System (INIS)

    Gilliam, T.M.; Jantzen, C.M.

    1996-10-01

    A model is presented based upon calculated bridging oxygens which allows the prediction of the region of acceptable glass compositions for a lime-soda-silica glass-forming system containing mixed waste. The model can be used to guide glass formulation studies (e.g., treatability studies) or assess the applicability of vitrification to candidate waste streams

  5. Vitrification of F006 plating waste sludge by Reactive Additive Stabilization Process (RASP)

    International Nuclear Information System (INIS)

    Martin, H.L.; Jantzen, C.M.; Pickett, J.B.

    1994-01-01

    Solidification into glass of nickel-on-uranium plating wastewater treatment plant sludge (F006 Mixed Waste) has been demonstrated at the Savannah River She (SRS). Vitrification using high surface area additives, the Reactive Additive Stabilization Process (RASP), greatly enhanced the solubility and retention of heavy metals In glass. The bench-scale tests using RASP achieved 76 wt% waste loading In both soda-lime-silica and borosilicate glasses. The RASP has been Independently verified by a commercial waste management company, and a contract awarded to vitrify the approximately 500,000 gallons of stored waste sludge. The waste volume reduction of 89% will greatly reduce the disposal costs, and delisting of the glass waste is anticipated. This will be the world's first commercial-scale vitrification system used for environmental cleanup of Mixed Waste. Its stabilization and volume reduction abilities are expected to set standards for the future of the waste management Industry

  6. Vitrification of high-level liquid wastes

    International Nuclear Information System (INIS)

    Varani, J.L.; Petraitis, E.J.; Vazquez, Antonio.

    1987-01-01

    High-level radioactive liquid wastes produced in the fuel elements reprocessing require, for their disposal, a preliminary treatment by which, through a series of engineering barriers, the dispersion into the biosphere is delayed by 10 000 years. Four groups of compounds are distinguished among a great variety of final products and methods of elaboration. From these, the borosilicate glasses were chosen. Vitrification experiences were made at a laboratory scale with simulated radioactive wastes, employing different compositions of borosilicate glass. The installations are described. A series of tests were carried out on four basic formulae using always the same methodology, consisting of a dry mixture of the vitreous matrix's products and a dry simulated mixture. Several quality tests of the glasses were made 1: Behaviour in leaching following the DIN 12 111 standard; 2: Mechanical resistance; parameters related with the facility of the different glasses for increasing their surface were studied; 3: Degree of devitrification: it is shown that devitrification turns the glasses containing radioactive wastes easily leachable. From all the glasses tested, the composition SiO 2 , Al 2 O 3 , B 2 O 3 , Na 2 O, CaO shows the best retention characteristics. (M.E.L.) [es

  7. Hanford Waste Vitrification Plant technical manual

    Energy Technology Data Exchange (ETDEWEB)

    Larson, D.E. [ed.; Watrous, R.A.; Kruger, O.L. [and others

    1996-03-01

    A key element of the Hanford waste management strategy is the construction of a new facility, the Hanford Waste Vitrification Plant (HWVP), to vitrify existing and future liquid high-level waste produced by defense activities at the Hanford Site. The HWVP mission is to vitrify pretreated waste in borosilicate glass, cast the glass into stainless steel canisters, and store the canisters at the Hanford Site until they are shipped to a federal geological repository. The HWVP Technical Manual (Manual) documents the technical bases of the current HWVP process and provides a physical description of the related equipment and the plant. The immediate purpose of the document is to provide the technical bases for preparation of project baseline documents that will be used to direct the Title 1 and Title 2 design by the A/E, Fluor. The content of the Manual is organized in the following manner. Chapter 1.0 contains the background and context within which the HWVP was designed. Chapter 2.0 describes the site, plant, equipment and supporting services and provides the context for application of the process information in the Manual. Chapter 3.0 provides plant feed and product requirements, which are primary process bases for plant operation. Chapter 4.0 summarizes the technology for each plant process. Chapter 5.0 describes the engineering principles for designing major types of HWVP equipment. Chapter 6.0 describes the general safety aspects of the plant and process to assist in safe and prudent facility operation. Chapter 7.0 includes a description of the waste form qualification program and data. Chapter 8.0 indicates the current status of quality assurance requirements for the Manual. The Appendices provide data that are too extensive to be placed in the main text, such as extensive tables and sets of figures. The Manual is a revision of the 1987 version.

  8. Hanford Waste Vitrification Plant technical manual

    International Nuclear Information System (INIS)

    Larson, D.E.; Watrous, R.A.; Kruger, O.L.

    1996-03-01

    A key element of the Hanford waste management strategy is the construction of a new facility, the Hanford Waste Vitrification Plant (HWVP), to vitrify existing and future liquid high-level waste produced by defense activities at the Hanford Site. The HWVP mission is to vitrify pretreated waste in borosilicate glass, cast the glass into stainless steel canisters, and store the canisters at the Hanford Site until they are shipped to a federal geological repository. The HWVP Technical Manual (Manual) documents the technical bases of the current HWVP process and provides a physical description of the related equipment and the plant. The immediate purpose of the document is to provide the technical bases for preparation of project baseline documents that will be used to direct the Title 1 and Title 2 design by the A/E, Fluor. The content of the Manual is organized in the following manner. Chapter 1.0 contains the background and context within which the HWVP was designed. Chapter 2.0 describes the site, plant, equipment and supporting services and provides the context for application of the process information in the Manual. Chapter 3.0 provides plant feed and product requirements, which are primary process bases for plant operation. Chapter 4.0 summarizes the technology for each plant process. Chapter 5.0 describes the engineering principles for designing major types of HWVP equipment. Chapter 6.0 describes the general safety aspects of the plant and process to assist in safe and prudent facility operation. Chapter 7.0 includes a description of the waste form qualification program and data. Chapter 8.0 indicates the current status of quality assurance requirements for the Manual. The Appendices provide data that are too extensive to be placed in the main text, such as extensive tables and sets of figures. The Manual is a revision of the 1987 version

  9. Radioactive waste vitrification offgas analysis proposal

    International Nuclear Information System (INIS)

    Nelson, C.W.; Morrey, E.V.

    1993-11-01

    Further validation of the Hanford Waste Vitrification Plant (HWVP) feed simulants will be performed by analyzing offgases during crucible melting of actual waste glasses and simulants. The existing method of vitrifying radioactive laboratory-scale samples will be modified to allow offgas analysis during preparation of glass for product testing. The analysis equipment will include two gas chromatographs (GC) with thermal conductivity detectors (TCD) and one NO/NO x analyzer. This equipment is part of the radioactive formating offgas system. The system will provide real-time analysis of H 2 , O 2 , N 2 , NO, N 2 O, NO 2 , CO, CO 2 , H 2 O, and SO 2 . As with the prior melting method, the product glass will be compatible with durability testing, i.e., Product Consistency Test (PCT) and Material Characterization Center (MCC-1), and crystallinity analysis. Procedures have been included to ensure glass homogeneity and quenching. The radioactive glass will be adaptable to Fe +2 /ΣFe measurement procedures because the atmosphere above the melt can be controlled. The 325 A-hot cell facility is being established as the permanent location for radioactive offgas analysis during formating, and can be easily adapted to crucible melt tests. The total costs necessary to set up and perform offgas measurements on the first radioactive core sample is estimated at $115K. Costs for repeating the test on each additional core sample are estimated to be $60K. The schedule allows for performing the test on the next available core sample

  10. Engineering-scale vitrification of commercial high-level waste

    International Nuclear Information System (INIS)

    Bonner, W.F.; Bjorklund, W.J.; Hanson, M.S.; Knowlton, D.E.

    1980-04-01

    To date, technology for immobilizing commercial high-level waste (HLW) has been extensively developed, and two major demonstration projects have been completed, the Waste Solidification Engineering Prototypes (WSEP) Program and the Nuclear Waste Vitrification Project (NWVP). The feasibility of radioactive waste solidification was demonstrated in the WSEP program between 1966 and 1970 (McElroy et al. 1972) using simulated power-reactor waste composed of nonradioactive chemicals and HLW from spent, Hanford reactor fuel. Thirty-three engineering-scale canisters of solidified HLW were produced during the operations. In early 79, the NWVP demonstrated the vitrification of HLW from the processing of actual commercial nuclear fuel. This program consisted of two parts, (1) waste preparation and (2) vitrification by spray calcination and in-can melting. This report presents results from the NWVP

  11. Characterization and vitrification of Hanford radioactive high level wastes

    International Nuclear Information System (INIS)

    Tingey, J.M.; Elliott, M.L.; Larson, D.E.; Morrey, E.V.

    1991-01-01

    Radioactive Neutralized Current Acid Waste (NCAW) samples from the Hanford waste tanks have been chemically, radiochemically and physically characterized. The wastes were processed according to the Hanford Waste vitrification Plant (HWVP) flowsheet, and characterized after each process step. The waste glasses were sectioned and leach tested. Chemical, radiochemical and physical properties of the waste will be presented and compared to nonradioactive simulant data and the HWVP reference composition and properties

  12. Design of microwave vitrification systems for radioactive waste

    International Nuclear Information System (INIS)

    White, T.L.; Wilson, C.T.; Schaick, C.R.; Bostick, W.D.

    1996-01-01

    Oak Ridge National Laboratory (ORNL) is involved in the research and development of high-power microwave heating systems for the vitrification of DOE radioactive sludges. Design criteria for a continuous microwave vitrification system capable of processing a surrogate filtercake sludge representative of a typical waste-water treatment operation are discussed. A prototype 915 MHz, 75 kW microwave vitrification system or 'microwave melter' is described along with some early experimental results that demonstrate a 4 to 1 volume reduction of a surrogate ORNL filtercake sludge

  13. Design of microwave vitrification systems for radioactive waste

    International Nuclear Information System (INIS)

    White, T.L.; Wilson, C.T.; Schaich, C.R.; Bostick, T.L.

    1995-01-01

    Oak Ridge National Laboratory (ORNL) is involved in the research and development of high-power microwave heating systems for the vitrification of Department of Energy (DOE) radioactive sludges. Design criteria for a continuous microwave vitrification system capable of processing a surrogate filtercake sludge representative of a typical waste-water treatment operation are discussed. A prototype 915-MHz, 75-kW microwave vitrification system or ''microwave melter'' is described along with some early experimental results that demonstrate a 4 to 1 volume reduction of a surrogate ORNL filtercake sludge

  14. High-level waste processing and conditioning: vitrification

    International Nuclear Information System (INIS)

    Bonniaud, R.

    1981-02-01

    The vitrification process used to treat fission product solutions at the Marcoule Vitrification Plant is described. The type of waste processed is characterized by its very high activity and the long lifetimes of some of the emitters that it contains. The performance obtained with this process is given together with the future developments envisaged. The storage of glasses is described as well as their behavior with time [fr

  15. In situ vitrification: application analysis for stabilization of transuranic waste

    International Nuclear Information System (INIS)

    Oma, K.H.; Farnsworth, R.K.; Rusin, J.M.

    1982-09-01

    The in situ vitrification process builds upon the electric melter technology previously developed for high-level waste immobilization. In situ vitrification converts buried wastes and contaminated soil to an extremely durable glass and crystalline waste form by melting the materials, in place, using joule heating. Once the waste materials have been solidified, the high integrity waste form should not cause future ground subsidence. Environmental transport of the waste due to water or wind erosion, and plant or animal intrusion, is minimized. Environmental studies are currently being conducted to determine whether additional stabilization is required for certain in-ground transuranic waste sites. An applications analysis has been performed to identify several in situ vitrification process limitations which may exist at transuranic waste sites. Based on the process limit analysis, in situ vitrification is well suited for solidification of most in-ground transuranic wastes. The process is best suited for liquid disposal sites. A site-specific performance analysis, based on safety, health, environmental, and economic assessments, will be required to determine for which sites in situ vitrification is an acceptable disposal technique. Process economics of in situ vitrification compare favorably with other in-situ solidification processes and are an order of magnitude less than the costs for exhumation and disposal in a repository. Leachability of the vitrified product compares closely with that of Pyrex glass and is significantly better than granite, marble, or bottle glass. Total release to the environment from a vitrified waste site is estimated to be less than 10 -5 parts per year. 32 figures, 30 tables

  16. Radioactive air emissions notice of construction and application for approval to construct the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    1992-10-01

    The Hanford Site is owned by the US Government and operated by the US Department of Energy, Richland Field Office. The Hanford Site manages and produces dangerous waste and mixed waste. (containing both radioactive and dangerous components). The US Department of Energy, Richland Field Office, currently stores mixed waste, resulting from various processing operations, in underground storage tanks. The Hanford Waste Vitrification Plant will be constructed and operated to process the high-activity fraction of mixed waste stored in these underground tanks. The Hanford Waste Vitrification Plant will solidify pretreated tank waste into a glass product that will be packaged for disposal in a national repository. Emissions from the Hanford Waste Vitrification Plant will be regulated by both the federal and state Clean Air Acts. The proposed Hanford Waste Vitrification Plant represents a new source of radioactive air emissions. Construction of the plant will require approval from both federal and state agencies. The Notice of Construction and Application for Approval to Construct the Hanford Waste Vitrification Plant contains information required under Title 40 of the Code of Federal Regulations, Chapter 61; and Chapter 246-247 of the Washington Administrative Code for a proposed new source of radioactive air emissions. The document contents are based on information contained in the Hanford Waste Vitrification Plant Reference Conceptual Design Report, the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report, Revision 0, and subsequent design changes made before August 1, 1992. The contents of this document may be modified to include more specific information generated during subsequent detailed design phases. Modifications will be submitted for regulatory review and approval, as appropriate

  17. Nuclear Waste Vitrification Efficiency: Cold Cap Reactions

    International Nuclear Information System (INIS)

    Kruger, A.A.; Hrma, P.R.; Pokorny, R.

    2011-01-01

    conditions. The model demonstrates that batch foaming has a decisive influence on the rate of melting. Understanding the dynamics of the foam layer at the bottom of the cold cap and the heat transfer through it appears crucial for a reliable prediction of the rate of melting as a function of the melter-feed makeup and melter operation parameters. Although the study is focused on a batch for waste vitrification, the authors expect that the outcome will also be relevant for commercial glass melting.

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

    International Nuclear Information System (INIS)

    Gaskill, J.R.; Larson, D.E.; Abrigo, G.P.

    1996-03-01

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

  19. Transportable Vitrification System RCRA Closure Practical Waste Disposition Saves Time And Money

    International Nuclear Information System (INIS)

    Brill, Angie; Boles, Roger; Byars, Woody

    2003-01-01

    The Transportable Vitrification System (TVS) was a large-scale vitrification system for the treatment of mixed wastes. The wastes contained both hazardous and radioactive materials in the form of sludge, soil, and ash. The TVS was developed to be moved to various United States Department of Energy (DOE) facilities to vitrify mixed waste as needed. The TVS consists of four primary modules: (1) Waste and Additive Materials Processing Module; (2) Melter Module; (3) Emissions Control Module; and (4) Control and Services Module. The TVS was demonstrated at the East Tennessee Technology Park (ETTP) during September and October of 1997. During this period, approximately 16,000 pounds of actual mixed waste was processed, producing over 17,000 pounds of glass. After the demonstration was complete it was determined that it was more expensive to use the TVS unit to treat and dispose of mixed waste than to direct bury this waste in Utah permitted facility. Thus, DOE had to perform a Resource Conservation and Recovery Act (RCRA) closure of the facility and find a reuse for as much of the equipment as possible. This paper will focus on the following items associated with this successful RCRA closure project: TVS site closure design and implementation; characterization activities focused on waste disposition; pollution prevention through reuse; waste minimization efforts to reduce mixed waste to be disposed; and lessons learned that would be integrated in future projects of this magnitude

  20. Feasibility Study for Vitrification of Sodium-Bearing Waste

    International Nuclear Information System (INIS)

    Quigley, J.J.; Raivo, B.D.; Bates, S.O.; Berry, S.M.; Nishioka, D.N.; Bunnell, P.J.

    2000-01-01

    Treatment of sodium-bearing waste (SBW) at the Idaho Nuclear Technology and Engineering Center (INTEC) within the Idaho National Engineering and Environmental Laboratory is mandated under a Settlement Agreement between the Department of Energy and the State of Idaho. One of the requirements of the Settlement Agreement is the complete calcination (i.e., treatment) of all SBW by December 31, 2012. One of the proposed options for treatment of SBW is vitrification. This study will examine the viability of SBW vitrification. This study describes the process and facilities to treat the SBW, from beginning waste input from INTEC Tank Farm to the final waste forms. Schedules and cost estimates for construction and operation of a Vitrification Facility are included. The study includes a facility layout with drawings, process description and flow diagrams, and preliminary equipment requirements and layouts

  1. Feasibility Study for Vitrification of Sodium-Bearing Waste

    Energy Technology Data Exchange (ETDEWEB)

    J. J. Quigley; B. D. Raivo; S. O. Bates; S. M. Berry; D. N. Nishioka; P. J. Bunnell

    2000-09-01

    Treatment of sodium-bearing waste (SBW) at the Idaho Nuclear Technology and Engineering Center (INTEC) within the Idaho National Engineering and Environmental Laboratory is mandated under a Settlement Agreement between the Department of Energy and the State of Idaho. One of the requirements of the Settlement Agreement is the complete calcination (i.e., treatment) of all SBW by December 31, 2012. One of the proposed options for treatment of SBW is vitrification. This study will examine the viability of SBW vitrification. This study describes the process and facilities to treat the SBW, from beginning waste input from INTEC Tank Farm to the final waste forms. Schedules and cost estimates for construction and operation of a Vitrification Facility are included. The study includes a facility layout with drawings, process description and flow diagrams, and preliminary equipment requirements and layouts.

  2. Crystalization and redox effects in waste vitrification

    International Nuclear Information System (INIS)

    Kim, C.W.; Buechele, A.C.; Muller, I.S.

    1996-01-01

    This is the continuation of a systematic study to determine the effects of redox state and the concentration of certain transition metals on selected properties of a simplified lime-aluminosilicate glass system, similar to one proposed for high temperature (1350 degrees C-1450 degrees C) vitrification of soil and wastes from DOE sites. The solubilities of Cr 2 O 3 , ZnO, NiO, and Fe 2 O 3 in the base glass, and of the first three oxides in higher-iron variants of the base glass are determined at 1350 degrees C, 1400 degrees C, and 1450 degrees C. Enthalpies of solution are calculated from the solubility data for these four transition metal oxides. Different redox ratios, Fe 2+ /Fe total , are induced at 1450 degrees C in a glass containing NiO at about 75% of its solubility limit at this temperature and related to changes in microstructure. A ZnO-SiO 2 -Fe 2 O 3 pseudoternary 1450 degrees C isotherm is determined and plotted over a wide range of compositions for glasses melted in air. Phases appearing are zincite-, hematite- and spinel-type phases. A Time-Temperature-Transformation (TTT) curve is plotted for a ZnO (12 wt%) containing glass using data from heat treatment studies, and the crystal layer growth rate of a melilite-type phase appearing in this glass is measured at several temperatures over the time range in which the rate is found to be linear. Some kinetic parameters of crystal growth are calculated

  3. The Hybrid Treatment Process for treatment of mixed radioactive and hazardous wastes

    International Nuclear Information System (INIS)

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

    1992-04-01

    This paper describes a new process for treating mixed hazardous and radioactive waste, commonly called mixed waste. The process is called the Hybrid Treatment Process (HTP), so named because it is built on the 20 years of experience with vitrification of wastes in melters, and the 12 years of experience with treatment of wastes by the in situ vitrification (ISV) process

  4. Seiler Pollution Control Systems vitrification process for the treatment of hazardous waste streams

    International Nuclear Information System (INIS)

    Nuesch, P.C.; Sarko, A.B.

    1995-01-01

    Seiler Pollution Control Systems, Inc. (Seiler) applies an economical, transportable, compact high temperature vitrification process to recycle and/or stabilize mixed organic/inorganic waste streams. Organic components are gasified by the system and are used as an auxiliary energy source. The inorganic components are melted and bound up molecularly in a glass/ceramic matrix. These glass/ceramics are extremely stable and durable and will pass typical regulatory leachate tests. Waste types that can be processed through the Seiler vitrification system include incinerator flyash, paint sludges, plating wastes, metal hydroxide sludges, low level and mixed radioactive wastes, contaminated soils and sludges, asbestos, and various mixed organic/inorganic residues. For nonradioactive waste streams, a variety of commercially saleable glass/ceramic products can be produced. These materials are marketed either as architectural materials, abrasives, or insulating refractories. The glass/ceramics generated from radioactive waste streams can be formed in a shape that is easily handled, stored, and retrieved. The system, itself is modular and can either be used as a stand alone system or hooked-up in line to existing manufacturing and production facilities. It consists of four sections: feed preparation; preheater; vitrifier/converter, and air pollution control. The vitrification system can use oxygen enriched natural gas or fuel oil for both cost efficiency and to reduce air pollution emissions

  5. Savannah River Site waste vitrification projects initiated throughout the United States: Disposal and recycle options

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    2000-01-01

    A vitrification process was developed and successfully implemented by the US Department of Energy's (DOE) Savannah River Site (SRS) and at the West Valley Nuclear Services (WVNS) to convert high-level liquid nuclear wastes (HLLW) to a solid borosilicate glass for safe long term geologic disposal. Over the last decade, SRS has successfully completed two additional vitrification projects to safely dispose of mixed low level wastes (MLLW) (radioactive and hazardous) at the SRS and at the Oak Ridge Reservation (ORR). The SRS, in conjunction with other laboratories, has also demonstrated that vitrification can be used to dispose of a wide variety of MLLW and low-level wastes (LLW) at the SRS, at ORR, at the Los Alamos National Laboratory (LANL), at Rocky Flats (RF), at the Fernald Environmental Management Project (FEMP), and at the Hanford Waste Vitrification Project (HWVP). The SRS, in conjunction with the Electric Power Research Institute and the National Atomic Energy Commission of Argentina (CNEA), have demonstrated that vitrification can also be used to safely dispose of ion-exchange (IEX) resins and sludges from commercial nuclear reactors. In addition, the SRS has successfully demonstrated that numerous wastes declared hazardous by the US Environmental Protection Agency (EPA) can be vitrified, e.g. mining industry wastes, contaminated harbor sludges, asbestos containing material (ACM), Pb-paint on army tanks and bridges. Once these EPA hazardous wastes are vitrified, the waste glass is rendered non-hazardous allowing these materials to be recycled as glassphalt (glass impregnated asphalt for roads and runways), roofing shingles, glasscrete (glass used as aggregate in concrete), or other uses. Glass is also being used as a medium to transport SRS americium (Am) and curium (Cm) to the Oak Ridge Reservation (ORR) for recycle in the ORR medical source program and use in smoke detectors at an estimated value of $1.5 billion to the general public

  6. Leaching characteristics of copper flotation waste before and after vitrification.

    Science.gov (United States)

    Coruh, Semra; Ergun, Osman Nuri

    2006-12-01

    Copper flotation waste from copper production using a pyrometallurgical process contains toxic metals such as Cu, Zn, Co and Pb. Because of the presence of trace amounts of these highly toxic metals, copper flotation waste contributes to environmental pollution. In this study, the leaching characteristics of copper flotation waste from the Black Sea Copper Works in Samsun, Turkey have been investigated before and after vitrification. Samples obtained from the factory were subjected to toxicity tests such as the extraction procedure toxicity test (EP Tox), the toxicity characteristic leaching procedure (TCLP) and the "method A" extraction procedure of the American Society of Testing and Materials. The leaching tests showed that the content of some elements in the waste before vitrification exceed the regulatory limits and cannot be disposed of in the present form. Therefore, a stabilization or inertization treatment is necessary prior to disposal. Vitrification was found to stabilize heavy metals in the copper flotation waste successfully and leaching of these metals was largely reduced. Therefore, vitrification can be an acceptable method for disposal of copper flotation waste.

  7. Independent engineering review of the Hanford Waste Vitrification System

    International Nuclear Information System (INIS)

    1991-10-01

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

  8. Independent engineering review of the Hanford Waste Vitrification System

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-01

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

  9. New developments for medium and low level waste vitrification

    International Nuclear Information System (INIS)

    Boen, A.J.-R.; Pujadas, S.M.-V.

    1997-01-01

    Converting ultimate waste material into a stable, inert product is beneficial, notably in the case of potentially very toxic wastes. Vitrification, in which a glass or glass-ceramic material is fabricated from a particular waste form, is now a proven solution. This high-temperature process uses additives-notably silica-if necessary to form a glass network. Vitrification confines the waste by forming a stable, inert, nontoxic material suitable for safe disposal; it usually also results in a significant volume reduction having a major effect on the disposal cost. France is actively engaged in an ongoing research effort in this area, not only to enhance the production capacity and the containment quality, but also to extend the process to low and medium level wastes such as those produced in nuclear power stations

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

    International Nuclear Information System (INIS)

    Hrma, P.R.

    1993-09-01

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

  11. Multipurpose optimization models for high level waste vitrification

    International Nuclear Information System (INIS)

    Hoza, M.

    1994-08-01

    Optimal Waste Loading (OWL) models have been developed as multipurpose tools for high-level waste studies for the Tank Waste Remediation Program at Hanford. Using nonlinear programming techniques, these models maximize the waste loading of the vitrified waste and optimize the glass formers composition such that the glass produced has the appropriate properties within the melter, and the resultant vitrified waste form meets the requirements for disposal. The OWL model can be used for a single waste stream or for blended streams. The models can determine optimal continuous blends or optimal discrete blends of a number of different wastes. The OWL models have been used to identify the most restrictive constraints, to evaluate prospective waste pretreatment methods, to formulate and evaluate blending strategies, and to determine the impacts of variability in the wastes. The OWL models will be used to aid in the design of frits and the maximize the waste in the glass for High-Level Waste (HLW) vitrification

  12. Structural and microstructural aspects of asbestos-cement waste vitrification

    Science.gov (United States)

    Iwaszko, Józef; Zawada, Anna; Przerada, Iwona; Lubas, Małgorzata

    2018-04-01

    The main goal of the work was to evaluate the vitrification process of asbestos-cement waste (ACW). A mixture of 50 wt% ACW and 50 wt% glass cullet was melted in an electric furnace at 1400 °C for 90 min and then cast into a steel mold. The vitrified product was subjected to annealing. Optical microscopy, scanning electron microscopy (SEM), Fourier transform infrared spectroscopy (FT-IR) and X-ray diffraction (XRD) were used to evaluate the effects of the vitrification. The chemical constitution of the material before and after the vitrification process was also analyzed. It was found that the vitrified product has an amorphous structure in which the components of asbestos-cement waste are incorporated. MIR spectroscopy showed that the absorption bands of chrysotile completely disappeared after the vitrification process. The results of the spectroscopic studies were confirmed by X-ray studies - no diffraction reflections from the chrysotile crystallographic planes were observed. As a result of the treatment, the fibrous asbestos construction, the main cause of its pathogenic properties, completely disappeared. The vitrified material was characterized by higher resistance to ion leaching in an aquatic environment than ACW and a smaller volume of nearly 72% in relation to the apparent volume of the substrates. The research has confirmed the high effectiveness of vitrification in neutralizing hazardous waste containing asbestos and the FT-IR spectroscopy was found to be useful to identify asbestos varieties and visualizing changes caused by the vitrification process. The work also presents the current situation regarding the utilization of asbestos-containing products.

  13. Glass melter assembly for the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Chen, A.E.; Russell, A.; Shah, K.R.; Kalia, J.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) is designed to solidify high level radioactive waste by converting it into stable borosilicate after mixing with glass frit and water. The heart of this conversion process takes place in the glass melter. The life span of the existing melter is limited by the possible premature failure of the heater assembly, which is not remotely replaceable, in the riser and pour spout. A goal of HWVP Project is to design remotely replaceable riser and pour spout heaters so that the useful life of the melter can be prolonged. The riser pour spout area is accessible only by the canyon crane and impact wrench. It is also congested with supporting frame members, service piping, electrode terminals, canister positioning arm and other various melter components. The visibility is low and the accessibility is limited. The problem is further compounded by the extreme high temperature in the riser core and the electrical conductive nature of the molten glass that flows through it

  14. Test Summary Report INEEL Sodium-Bearing Waste Vitrification Demonstration RSM-01-1

    Energy Technology Data Exchange (ETDEWEB)

    Goles, Ronald W.; Perez, Joseph M.; Macisaac, Brett D.; Siemer, Darryl D.; Mccray, John A.

    2001-05-21

    The U.S. Department of Energy's Idaho National Engineering and Environmental Laboratory is storing large amounts of radioactive and mixed wastes. Most of the sodium-bearing wastes have been calcined, but about a million gallons remain uncalcined, and this waste does not meet current regulatory requirements for long-term storage and/or disposal. As a part of the Settlement Agreement between DOE and the State of Idaho, the tanks currently containing SBW are to be taken out of service by December 31, 2012, which requires removing and treatment the remaining SBW. Vitrification is the option for waste disposal that received the highest weighted score against the criteria used. Beginning in FY 2000, the INEEL high-level waste program embarked on a program for technology demonstration and development that would lead to conceptual design of a vitrification facility in the event that vitrification is the preferred alternative for SBW disposal. The Pacific Northwest National Laborator's Research-Scale Melter was used to conduct these initial melter-flowsheet evaluations. Efforts are underway to reduce the volume of waste vitrified, and during the current test, an overall SBW waste volume-reduction factor of 7.6 was achieved.

  15. Glass formulation for phase 1 high-level waste vitrification

    International Nuclear Information System (INIS)

    Vienna, J.D.; Hrma, P.R.

    1996-04-01

    The purpose of this study is to provide potential glass formulations for prospective Phase 1 High-Level Waste (HLW) vitrification at Hanford. The results reported here will be used to aid in developing a Phase 1 HLW vitrification request for proposal (RFP) and facilitate the evaluation of ensuing proposals. The following factors were considered in the glass formulation effort: impact on total glass volume of requiring the vendor to process each of the tank compositions independently versus as a blend; effects of imposing typical values of B 2 O 3 content and waste loading in HLW borosilicate glasses as restrictions on the vendors (according to WAPS 1995, the typical values are 5--10 wt% B 2 O 3 and 20--40 wt% waste oxide loading); impacts of restricting the processing temperature to 1,150 C on eventual glass volume; and effects of caustic washing on any of the selected tank wastes relative to glass volume

  16. Material chemistry challenges in vitrification of high level radioactive waste

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2008-01-01

    Full text: Nuclear technology with an affective environmental management plan and focused attention on safety measures is a much cleaner source of electricity generation as compared to other sources. With this perspective, India has undertaken nuclear energy program to share substantial part of future need of power. Safe containment and isolation of nuclear waste from human environment is an indispensable part of this programme. Majority of radioactivity in the entire nuclear fuel cycle is high level radioactive liquid waste (HLW), which is getting generated during reprocessing of spent nuclear fuels. A three stage strategy for management of HLW has been adopted in India. This involves (i) immobilization of waste oxides in stable and inert solid matrices, (ii) interim retrievable storage of the conditioned waste product under continuous cooling and (iii) disposal in deep geological formation. Borosilicate glass matrix has been adopted in India for immobilization of HLW. Material issue are very important during the entire process of waste immobilization. Performance of the materials used in nuclear waste management determines its safety/hazards. Material chemistry therefore has a significant bearing on immobilization science and its technological development for management of HLW. The choice of suitable waste form to deploy for nuclear waste immobilization is difficult decision and the durability of the conditioned product is not the sole criterion. In any immobilization process, where radioactive materials are involved, the process and operational conditions play an important role in final selection of a suitable glass formulation. In remotely operated vitrification process, study of chemistry of materials like glass, melter, materials of construction of other equipment under high temperature and hostile corrosive condition assume significance for safe and un-interrupted vitrification of radioactive to ensure its isolation waste from human environment. The present

  17. Selection of melter systems for the DOE/Industrial Center for Waste Vitrification Research

    International Nuclear Information System (INIS)

    Bickford, D.F.

    1993-01-01

    The EPA has designated vitrification as the best developed available technology for immobilization of High-Level Nuclear Waste. In a recent federal facilities compliance agreement between the EPA, the State of Washington, and the DOE, the DOE agreed to vitrify all of the Low Level Radioactive Waste resulting from processing of High Level Radioactive Waste stored at the Hanford Site. This is expected to result in the requirement of 100 ton per day Low Level Radioactive Waste melters. Thus, there is increased need for the rapid adaptation of commercial melter equipment to DOE's needs. DOE has needed a facility where commercial pilot scale equipment could be operated on surrogate (non-radioactive) simulations of typical DOE waste streams. The DOE/Industry Center for Vitrification Research (Center) was established in 1992 at the Clemson University Department of Environmental Systems Engineering, Clemson, SC, to address that need. This report discusses some of the characteristics of the melter types selected for installation of the Center. An overall objective of the Center has been to provide the broadest possible treatment capability with the minimum number of melter units. Thus, units have been sought which have broad potential application, and which had construction characteristics which would allow their adaptation to various waste compositions, and various operating conditions, including extreme variations in throughput, and widely differing radiological control requirements. The report discusses waste types suitable for vitrification; technical requirements for the application of vitrification to low level mixed wastes; available melters and systems; and selection of melter systems. An annotated bibliography is included

  18. Selection of melter systems for the DOE/Industrial Center for Waste Vitrification Research

    Energy Technology Data Exchange (ETDEWEB)

    Bickford, D.F.

    1993-12-31

    The EPA has designated vitrification as the best developed available technology for immobilization of High-Level Nuclear Waste. In a recent federal facilities compliance agreement between the EPA, the State of Washington, and the DOE, the DOE agreed to vitrify all of the Low Level Radioactive Waste resulting from processing of High Level Radioactive Waste stored at the Hanford Site. This is expected to result in the requirement of 100 ton per day Low Level Radioactive Waste melters. Thus, there is increased need for the rapid adaptation of commercial melter equipment to DOE`s needs. DOE has needed a facility where commercial pilot scale equipment could be operated on surrogate (non-radioactive) simulations of typical DOE waste streams. The DOE/Industry Center for Vitrification Research (Center) was established in 1992 at the Clemson University Department of Environmental Systems Engineering, Clemson, SC, to address that need. This report discusses some of the characteristics of the melter types selected for installation of the Center. An overall objective of the Center has been to provide the broadest possible treatment capability with the minimum number of melter units. Thus, units have been sought which have broad potential application, and which had construction characteristics which would allow their adaptation to various waste compositions, and various operating conditions, including extreme variations in throughput, and widely differing radiological control requirements. The report discusses waste types suitable for vitrification; technical requirements for the application of vitrification to low level mixed wastes; available melters and systems; and selection of melter systems. An annotated bibliography is included.

  19. Vitrification of transuranic and beta-gamma contaminated solid wastes

    International Nuclear Information System (INIS)

    Dukes, M.D.

    1980-06-01

    Vitrification of solid transuranic contaminated (TRU) wastes alone and with high-level liquid wastes (HLLW) was studied. Homogeneous glasses containing 20 to 30 wt % ash were made by using glass frits previously developed at the Savannah River Plant and Pacific Northwest Laboratories. If the ash is vitrified along with the HLLW, 1.0 wt % as can be added to the waste forms without affecting their quality. This loading of ash is well above the loading required by the relative amounts of HLLW and TRU ash that will be processed at the Savannah River Plant. Vitrification of TRU-contaminated electropolishing sludges and high efficiency particular air filter materials along with HLLW would require an increase in the quantity of glass to be produced. However, if these TRU-contaminated solids were vitrified with the HLLW, the addition of low-level beta-gamma contaminated ash would require no further increase in glass production

  20. Hanford Waste Vitrification Plant dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-10-01

    This report presents engineering drawings of the vitrification plant at Hanford Reservation. Individual sections in the report cover piping and instrumentation, process flow schemes, and material balance tables

  1. Vitrification of high-level alumina nuclear waste

    International Nuclear Information System (INIS)

    Brotzman, J.R.

    1979-01-01

    Borophosphate glass compositions have been developed for the vitrification of a high-alumina calcined defense waste. The effect of substituting SiO 2 , P 2 O 5 and CuO for B 2 O 3 on the viscosity and leach resistance was measured. The effect of the alkali to borate ratio and the Li 2 O:Na 2 O ratio on the melt viscosity and leach resistance was also measured

  2. Hanford Waste Vitrification Plant: Preliminary description of waste form and canister

    International Nuclear Information System (INIS)

    Mitchell, D.E.

    1986-01-01

    In July 1985, the US Department of Energy's Office of Civilian Radioactive Waste Management established the Waste Acceptance Process as the means by which defense high-level waste producers, such as the Hanford Waste Vitrification Plant, will develop waste acceptance requirements with the candidate geologic repositories. A complete description of the Waste Acceptance Process is contained in the Preliminary Hanford Waste Vitrification Plant Waste Form Qualification Plan. The Waste Acceptance Process defines three documents that high-level waste producers must prepare as a part of the process of assuming that a high-level waste product will be acceptable for disposal in a geologic repository. These documents are the Description of Waste Form and Canister, Waste Compliance Plan, and Waste Qualification Report. This document is the Hanford Waste Vitrification Plant Preliminary Description of Waste Form and Canister for disposal of Neutralized Current Acid Waste. The Waste Acceptance Specifications for the Hanford Waste Vitrification Plant have not yet been developed, therefore, this document has been structured to corresponds to the Waste Acceptance Preliminary Specifications for the Defense Waste Processing Facility High-Level Waste Form. Not all of the information required by these specifications is appropriate for inclusion in this Preliminary Description of Waste Form and Canister. Rather, this description is limited to information that describes the physical and chemical characteristics of the expected high-level waste form. The content of the document covers three major areas: waste form characteristics, canister characteristics, and canistered waste form characteristics. This information will be used by the candidate geologic repository projects as the basis for preliminary repository design activities and waste form testing. Periodic revisions are expected as the Waste Acceptance Process progresses

  3. Upfront Delisting of F006 Mixed Waste

    International Nuclear Information System (INIS)

    Poulos, D.G.; Pickett, J.B.; Jantzen, C.M.

    1995-01-01

    The US DOE at the Savannah River Site will petition the US EPA to upfront delist treatment residues generated from the vitrification of approximately 650,000 gallons of a regulated mixed (hazardous and radioactive) waste. The upfront petition, based on bench-scale treatability studies and pilot-scale system data, will exclude the vitrified wasteform from hazardous waste management regulations. The EPA encourages the use of the upfront delisting method as it allows applicants prior knowledge of waste specific treatment standards, which when met will render the waste non-hazardous, before generating the final wasteform. To meet the EPA performance based treatment standards, the waste must be stabilized to control the leaching of hazardous and radioactive constituents from the final wasteform. SRS has contracted a vendor to stabilize the mixed waste in a temporary Vitrification Treatment Facility (VTF). The EPA has declared vitrification as the Best Demonstrated Available Technology for high level radioactive wastes and the DOE Office of Technology Development has taken the position that mixed waste needs to be stabilized to the highest degree possible to ensure that the resulting wasteform meets both current and future regulatory specifications. Treatability studies conducted on a VTF pilot-scale system unit indicates that the mixed waste can be converted into a highly durable glass form, which exceeds the projected EPA performance based criteria. Upfront petitions can be processed by the EPA concurrently during facility construction or permitting activities; therefore, the SRS VTF will be capable of producing wastes which are considered non-hazardous sooner than otherwise expected. At the same time, EPA imposed conditional testing requirements to verify that the delisting levels are achieved by the fully operational VTF, ensures that only non-hazardous wastes are removed from hazardous waste management regulations. Vitrification of the (Abstract Truncated)

  4. The role of frit in nuclear waste vitrification

    International Nuclear Information System (INIS)

    Vienna, J.D.; Smith, P.A.; Dorn, D.A.; Hrma, P.

    1994-04-01

    Vitrification of nuclear waste requires additives which are often vitrified independently to form a frit. Frit composition is formulated to meet the needs of glass composition and processing. The effects of frit on melter feed and melt processing, glass acceptance, and waste loading is of practical interest in understanding the trade-offs associated with the competing demands placed on frit composition. Melter feed yield stress, viscosity and durability of frits and corresponding waste glasses as well as the kinetics of elementary melting processes have been measured. The results illustrate the competing requirements on frit. Four frits (FY91, FY93, HW39-4, and SR202) and simulated neutralized current acid waste (NCAW) were used in this study. The experimental evidence shows that optimization of frit for one processing related property often results in poorer performance for the remaining properties. The difficulties associated with maximum waste loading and durability are elucidated for glasses which could be processed using technology available for the previously proposed Hanford Waste Vitrification Plant

  5. The vitrification of high level wastes using microwave power

    International Nuclear Information System (INIS)

    Hardwick, W.H.; Gayler, R.; Murphy, V.

    1981-01-01

    A process for radioactive waste vitrification which exploits advantages peculiar to microwave heating is under development. The advantages claimed are the removal of the heat source from the radioactive environment, the elimination of heat transfer barriers by direct coupling of the energy with the process materials, and the ability to evaporate liquors absorbed in a glass fibre matrix which constitutes the glass forming additive. This glass fibre matrix which constitutes the glass forming additive. This glass fibre is also used to filter off-gases and give a condensate free of solids. The fibre loaded with dried waste is converted to a homogeneous glass by melting using microwave power. (orig./DG)

  6. Hanford Waste Vitrification Project Building limited scope risk assessment

    International Nuclear Information System (INIS)

    Braun, D.J.; Lindberg, S.E.; Reardon, M.F.; Wilson, G.P.

    1992-10-01

    A limited scope risk assessment was performed on the preliminary design of a high-level waste interim storage facility. The Canister Storage Building (CSB) facility will be built to support remediation at the US Department of Energy Hanford Site in Washington State. The CSB will be part of the support facilities for a high level Hanford Waste Vitrification Plant (HWVP). The limited scope risk assessment is based on a preliminary design which uses forced air circulation systems to move air through the building vault. The current building design calls for natural circulation to move air through the building vault

  7. First use of in situ vitrification on radioactive wastes

    International Nuclear Information System (INIS)

    Bowlds, L.

    1992-01-01

    A high-temperature method for containing hazardous wastes, which was first developed in the 1980s, is being adapted for the in situ treatment of buried radioactive wastes by the US DOE's Idaho National Engineering Laboratory (INEL), following its recent report on successful preliminary tests. The method, called in situ vitrification (ISV), is an electrically induced thermal process that melts and fuses soil and wastes into a glass-like material at least as strong as natural obsidian or granite. Gases released during the process are captured and treated by an off-gas treatment system. After the wastes are vitrified, they could be left in place, or the mass could be broken up and transported to a disposal site. The glass-like substance would be chemically and physically similar to obsidian and from 4 to 10 times more durable than typical borosilicate glasses used to immobolize high-level nuclear wastes

  8. Hanford Waste Vitrification Plant - the project and process systems

    International Nuclear Information System (INIS)

    Swenson, L.D.; Miller, W.C.; Smith, R.A.

    1990-01-01

    The Hanford Waste Vitrification Plant (HWVP) project is scheduled to start construction on the Hanford reservation in southeastern Washington State in 1991. The project will immobilize the liquid high-level defense waste stored there. The HWVP represents the third phase of the U.S. Department of Energy (DOE) activities that are focused on the permanent disposal of high-level radioactive waste, building on the experience of Defense Waste Processing Facility (DWPF) at the Savannah River site, South Carolina, and of the West Valley Demonstration Plant (WVDP), New York. This sequential approach to disposal of the country's commercial and defense high-level radioactive waste allows HWVP to make extensive use of lessons learned from the experience of its predecessors, using mature designs from the earlier facilities to achieve economies in design and construction costs while enhancing operational effectiveness

  9. Vitrification of TRU wastes at Rocky Flats Plant

    International Nuclear Information System (INIS)

    Williams, P.M.; Johnson, A.J.; Ledford, J.A.

    1979-01-01

    Immobilization of incinerator ash and various noncombustible TRU wastes was investigated. In three different research projects borosilicate glass proved to be the best candidate for TRU waste fixation. This glass has excellent chemical durability, long-term stability in the presence of radiation, and will withstand continuous temperatures up to 400 0 C without devitrification. In addition, wastes prepared in the form of glass will attain densities of approximately 2500 kg/m 3 (2.5 g/cc). The free forming method of producing glass buttons provides a very simple, consistent, low maintenance way of producing a final waste form for transporting and either retrievable or permanent storage for TRU waste. The vitrification process produces a durable glass from the low density ash generated by the fluidized bed incinerator process and provides volume and weight reductions that are superior to other fixation processes. This results in decreased transportation and storage costs

  10. Defense-waste vitrification studies during FY-1981. Summary report

    International Nuclear Information System (INIS)

    Bjorklund, W.J.

    1982-09-01

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

  11. Vitrification of Simulated Fernald K-65 Silo Waste at Low Temperature

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Pickett, J.B.

    1998-01-01

    degrees C. The VPP began operation on June 19, 1996. The VPP was used to test surrogate FEMP wastes at melt temperatures between 1130 degrees C and 1350 degrees C. The VPP failed on December 26, 1996 while processing surrogate waste. After the failure of the FEMP VPP, vitrification technology and glass chemistry were reevaluated. This report documents the glass formulation development for K-65 waste completed at SRTC in April, 1993 in conjunction with Associated Technical Consultants (ATC) of Toledo, Ohio. The glass developed for the FEMP was formulated in a lithia substituted soda-lime-silica (SLS) glass per the Savannah River Technology Center (SRTC) patented Lithia Additive Melting Process (LAMP)* to avoid problematic phase separation known to occur in the borosilicate glass system (MO-B 2 O 3 - SiO 2 ), where (MO = CaO, MgO, BaO, and PbO). Lime, MgO, BaO and PbO are all constituents of the FEMP wastes and thus subject to phase separation when vitrified in borosilicate glass. Phase separation is known to compromise waste glass stability. The SRTC soda-lithia-lime- silica (SLLS) glass melted at 1050 degrees C. Similar SLLS glass formulations have recently been demonstrated at the Oak Ridge Reservation (ORR) in a full scale melter with mixed (radioactive and hazardous) wastes.The low melting temperatures achieved with the SLLS glass minimize volatilization of hazardous species such as arsenic, lead, and selenium during vitrification. An 81 percent K-65 waste loading was demonstra

  12. Glass formulation for phase 1 high-level waste vitrification

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, J.D.; Hrma, P.R.

    1996-04-01

    The purpose of this study is to provide potential glass formulations for prospective Phase 1 High-Level Waste (HLW) vitrification at Hanford. The results reported here will be used to aid in developing a Phase 1 HLW vitrification request for proposal (RFP) and facilitate the evaluation of ensuing proposals. The following factors were considered in the glass formulation effort: impact on total glass volume of requiring the vendor to process each of the tank compositions independently versus as a blend; effects of imposing typical values of B{sub 2}O{sub 3} content and waste loading in HLW borosilicate glasses as restrictions on the vendors (according to WAPS 1995, the typical values are 5--10 wt% B{sub 2}O{sub 3} and 20--40 wt% waste oxide loading); impacts of restricting the processing temperature to 1,150 C on eventual glass volume; and effects of caustic washing on any of the selected tank wastes relative to glass volume.

  13. Corrosion assessment of refractory materials for high temperature waste vitrification

    International Nuclear Information System (INIS)

    Marra, J.C.; Congdon, J.W.; Kielpinski, A.L.

    1995-01-01

    A variety of vitrification technologies are being evaluated to immobilize radioactive and hazardous wastes following years of nuclear materials production throughout the Department of Energy (DOE) complex. The compositions and physical forms of these wastes are diverse ranging from inorganic sludges to organic liquids to heterogeneous debris. Melt and off-gas products can be very corrosive at the high temperatures required to melt many of these waste streams. Ensuring material durability is required to develop viable treatment processes. Corrosion testing of materials in some of the anticipated severe environments is an important aspect of the materials identification and selection process. Corrosion coupon tests on typical materials used in Joule heated melters were completed using glass compositions with high salt contents. The presence of chloride in the melts caused the most severe attack. In the metal alloys, oxidation was the predominant corrosion mechanism, while in the tested refractory material enhanced dissolution of the refractory into the glass was observed. Corrosion testing of numerous different refractory materials was performed in a plasma vitrification system using a surrogate heterogeneous debris waste. Extensive corrosion was observed in all tested materials

  14. Quality assurance program description: Hanford Waste Vitrification Plant, Part 1

    International Nuclear Information System (INIS)

    1992-01-01

    This document describes the Department of Energy's Richland Field Office (DOE-RL) quality assurance (QA) program for the processing of high-level waste as well as the Vitrification Project Quality Assurance Program for the design and construction of the Hanford Waste Vitrification Plant (HWVP). It also identifies and describes the planned activities that constitute the required quality assurance program for the HWVP. This program applies to the broad scope of quality-affecting activities associated with the overall HWVP Facility. Quality-affecting activities include designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, maintaining, repairing, and modifying. Also included are the development, qualification, and production of waste forms which may be safely used to dispose of high-level radioactive waste resulting from national defense activities. The HWVP QA program is made up of many constituent programs that are being implemented by the participating organizations. This Quality Assurance program description is intended to outline and define the scope and application of the major programs that make up the HWVP QA program. It provides a means by which the overall program can be managed and directed to achieve its objectives. Subsequent parts of this description will identify the program's objectives, its scope, application, and structure

  15. Vitrification of liquid waste from nuclear power plants

    International Nuclear Information System (INIS)

    Sheng Jiawei; Choi, Kwansik; Song, Myung-Jae

    2001-01-01

    Glass is an acceptable waste form to solidify the low-level waste from nuclear power plants (NPPs) because of the simplicity of processing and its unique ability to accept a wide variety of waste streams. Vitrification is being considered to solidify the high-boron-containing liquid waste generated from Korean NPPs. This study dealt with the development of a glass formulation to solidify the liquid waste. Studies were conducted in a borosilicate glass system. Crucible studies have been performed with surrogate waste. Several developed glass frits were evaluated to determine their suitability for vitrifying the liquid waste. The results indicated that the 20 wt% waste oxides loading required could not be obtained using these glass frits. Flyash produced from coal-burning electric power stations, whose major components are SiO 2 and Al 2 O 3 , is a desirable glass network former. Detailed product evaluations including waste loading, homogeneity, chemical durability and viscosity, etc., were carried out on selected formulations using flyash. Up to 30 wt% of the waste oxides was successfully solidified into the flyash after the addition of 5-10 wt% Na 2 O at 1200 deg. C

  16. Performance Enhancements to the Hanford Waste Treatment and Immobilization Plant Low-Activity Waste Vitrification System

    International Nuclear Information System (INIS)

    Hamel, W. F.; Gerdes, K.; Holton, L. K.; Pegg, I.L.; Bowan, B.W.

    2006-01-01

    The U.S Department of Energy Office of River Protection (DOE-ORP) is constructing a Waste Treatment and Immobilization Plant (WTP) for the treatment and vitrification of underground tank wastes stored at the Hanford Site in Washington State. The WTP comprises four major facilities: a pretreatment facility to separate the tank waste into high level waste (HLW) and low-activity waste (LAW) process streams, a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction, and an analytical laboratory to support the operations of all four treatment facilities. DOE has established strategic objectives to optimize the performance of the WTP facilities and the LAW and HLW waste forms to reduce the overall schedule and cost for treatment and vitrification of the Hanford tank wastes. This strategy has been implemented by establishing performance expectations in the WTP contract for the facilities and waste forms. In addition, DOE, as owner-operator of the WTP facilities, continues to evaluate 1) the design, to determine the potential for performance above the requirements specified in the WTP contract; and 2) improvements in production of the LAW and HLW waste forms. This paper reports recent progress directed at improving production of the LAW waste form. DOE's initial assessment, which is based on the work reported in this paper, is that the treatment rate of the WTP LAW vitrification facility can be increased by a factor of 2 to 4 with a combination of revised glass formulations, modest increases in melter glass operating temperatures, and a second-generation LAW melter with a larger surface area. Implementing these improvements in the LAW waste immobilization capability can benefit the LAW treatment mission by reducing the cost of waste treatment. (authors)

  17. Behavior of technetium in nuclear waste vitrification processes.

    Science.gov (United States)

    Pegg, Ian L

    Nearly 100 tests were performed with prototypical melters and off-gas system components to investigate the extents to which technetium is incorporated into the glass melt, partitioned to the off-gas stream, and captured by the off-gas treatment system components during waste vitrification. The tests employed several simulants, spiked with 99m Tc and Re (a potential surrogate), of the low activity waste separated from nuclear wastes in storage in the Hanford tanks, which is planned for immobilization in borosilicate glass. Single-pass technetium retention averaged about 35 % and increased significantly with recycle of the off-gas treatment fluids. The fraction escaping the recycle loop was very small.

  18. Savannah River Site chemical, metal, and pesticide (CMP) waste vitrification treatability studies

    International Nuclear Information System (INIS)

    Cicero, C.A.

    1997-01-01

    Numerous Department of Energy (DOE) facilities, as well as Department of Defense (DOD) and commercial facilities, have used earthen pits for disposal of chemicals, organic contaminants, and other waste materials. Although this was an acceptable means of disposal in the past, direct disposal into earthen pits without liners or barriers is no longer a standard practice. At the Savannah River Site (SRS), approximately three million pounds of such material was removed from seven chemical, metal, and pesticide disposal pits. This material is known as the Chemical, Metal, and Pesticide (CMP) Pit waste and carries several different listed waste codes depending on the contaminants in the respective storage container. The waste is not classified as a mixed waste because it is believed to be non-radioactive; however, in order to treat the material in a non-radioactive facility, the waste would first have to be screened for radioactivity. The Defense Waste Processing Technology (DWPT) Section of the Savannah River Technology Center (SRTC) was requested by the DOE-Savannah River (SR) office to determine the viability of vitrification of the CMP Pit wastes. Radioactive vitrification facilities exist which would be able to process this waste, so the material would not have to be analyzed for radioactive content. Bench-scale treatability studies were performed by the DWPT to determine whether a homogeneous and durable glass could be produced from the CMP Pit wastes. Homogeneous and durable glasses were produced from the six pits sampled. The optimum composition was determined to be 68.5 wt% CMP waste, 7.2 wt% Na 2 O, 9 wt% CaO, 7.2 wt% Li 2 O and 8.1 wt% Fe 2 O 3 . This glass melted at 1,150 C and represented a two fold volume reduction

  19. The Radiation Effect to Waste Glass that Resulting of Vitrification

    International Nuclear Information System (INIS)

    Herlan Martono; Aisyah

    2002-01-01

    The high level liquid waste (HLLW) is generated from the first step extraction of the nuclear fuel reprocessing. This waste was contain of few of actinide and many of fission product. The alpha radiation of actinide that contain on the HLLW cause the change the waste glass characteristic. The experiment was conducted by the doping, irradiation and heating of waste glass resulting from vitrification. The alpha radiation cause the change of composition that could be detected from change of waste glass density and mechanical strength. The increasing of alpha radiation dose cause the increasing change of density and mechanical strength, although the change of mechanical strength is not significant. Degree of change of waste glass density also depend on type of waste-glass and reach for saturated point at over of 5x10 24 alpha decay/m 3 . The gamma radiation of fission product that contain on the HLLW can increasing of waste glass temperature that cause the structure change, so devitrification was occur. The devitrification can the increasing of leaching rate. The cumulative of gamma dose rate was not cause the devitrification. (author)

  20. Chemical durability of soda-lime-aluminosilicate glass for radioactive waste vitrification

    International Nuclear Information System (INIS)

    Eppler, F.H.; Yim, M.S.

    1998-01-01

    Vitrification has been identified as one of the most viable waste treatment alternatives for nuclear waste disposal. Currently, the most popular glass compositions being selected for vitrification are the borosilicate family of glasses. Another popular type that has been around in glass industry is the soda-lime-silicate variety, which has often been characterized as the least durable and a poor candidate for radioactive waste vitrification. By replacing the boron constituent with a cheaper substitute, such as silica, the cost of vitrification processing can be reduced. At the same time, addition of network intermediates such as Al 2 O 3 to the glass composition increases the environmental durability of the glass. The objective of this study is to examine the ability of the soda-lime-aluminosilicate glass as an alternative vitrification tool for the disposal of radioactive waste and to investigate the sensitivity of product chemical durability to variations in composition

  1. High-level waste vitrification off-gas cleanup technology

    International Nuclear Information System (INIS)

    Hanson, M.S.

    1980-01-01

    This brief overview is intended to be a basis for discussion of needs and problems existing in the off-gas clean-up technology. A variety of types of waste form and processes are being developed in the United States and abroad. A description of many of the processes can be found in the Technical Alternative Documents (TAD). Concurrently, off-gas processing systems are being developed with most of the processes. An extensive review of methodology as well as decontamination factors can be found in the literature. Since it is generally agreed that the most advanced solidification process is vitrification, discussion here centers about the off-gas problems related to vitrification. With a number of waste soldification facilities around the world in operation, it can be shown that present technology can satisfy the present requirement for off-gas control. However, a number of areas within the technology base show potential for improvement. Fundamental as well as verification studies are needed to obtain the improvements

  2. High level radioactive waste vitrification process equipment component testing

    International Nuclear Information System (INIS)

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

    1985-04-01

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

  3. Low-level radioactive waste vitrification: effect of Cs partitioning

    International Nuclear Information System (INIS)

    Horton, W.S.; Ougouag, A.M.

    1986-01-01

    The traditional Low-Level Radioactive Waste (LLW) immobilization options are cementation or bituminization. Either of these options could be followed by shallow-land burial (SLB) or above-ground disposal. These rather simple LLW procedures appeared to be readily available, to meet regulatory requirements, and to satisfy cost constraints. The authorization of State Compacts, the forced closure of half of the six SLB disposal facilities of the nation, and the escalation of transportation/disposal fees diminish the viability of these options. The synergetic combination of these factors led to a reassessment of traditional methods and to an investigation of other techniques. This paper analyzes the traditional LLW immobilization options, reviews the impact of the LLW stream composition on Low-Level Waste Vitrification (LLWV), then proposes and briefly discusses several techniques to control the volatile radionuclides in a Process Improved LLWV system (PILLWV)

  4. Review of FY 2001 Development Work for Vitrification of Sodium Bearing Waste

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Dean Dalton; Barnes, Charles Marshall

    2002-09-01

    Treatment of sodium-bearing waste (SBW) at the Idaho Nuclear Technology and Engineering Center (INTEC) within the Idaho National Engineering and Environmental Laboratory is mandated by the Settlement Agreement between the Department of Energy and the State of Idaho. This report discusses significant findings from vitrification technology development during 2001 and their impacts on the design basis for SBW vitrification.

  5. Review of FY2001 Development Work for Vitrification of Sodium Bearing Waste

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, C.M.; Taylor, D.D.

    2002-09-09

    Treatment of sodium-bearing waste (SBW) at the Idaho Nuclear Technology and Engineering Center (INTEC) within the Idaho National Engineering and Environmental Laboratory is mandated by the Settlement Agreement between the Department of Energy and the State of Idaho. This report discusses significant findings from vitrification technology development during 2001 and their impacts on the design basis for SBW vitrification.

  6. Design of equipment used for high-level waste vitrification at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Vance, R.F.; Brill, B.A.; Carl, D.E.

    1997-06-01

    The equipment as designed, started, and operated for high-level radioactive waste vitrification at the West Valley Demonstration Project in western New York State is described. Equipment for the processes of melter feed make-up, vitrification, canister handling, and off-gas treatment are included. For each item of equipment the functional requirements, process description, and hardware descriptions are presented

  7. Review of FY2001 Development Work for Vitrification of Sodium Bearing Waste

    International Nuclear Information System (INIS)

    Barnes, C.M.; Taylor, D.D.

    2002-01-01

    Treatment of sodium-bearing waste (SBW) at the Idaho Nuclear Technology and Engineering Center (INTEC) within the Idaho National Engineering and Environmental Laboratory is mandated by the Settlement Agreement between the Department of Energy and the State of Idaho. This report discusses significant findings from vitrification technology development during 2001 and their impacts on the design basis for SBW vitrification

  8. Behavior of mercury and iodine during vitrification of simulated alkaline Purex waste

    International Nuclear Information System (INIS)

    Holton, L.K.

    1981-09-01

    Current plans indicate that the high-level wastes stored at the Savannah River Plant will be solidified by vitrification. The behavior of mercury and iodine during the vitrification process is of concern because: mercury is present in the waste in high concentrations (0.1 to 2.8 wt%); mercury will react with iodine and the other halogens present in the waste during vitrification and; the mercury compounds formed will be volatilized from the vitrification process placing a high particulate load in the vitrification system off-gas. Twelve experiments were completed to study the behavior of mercury during vitrification of simulated SRP Purex waste. The mercury was completely volatized from the vitrification system in all experiments. The mercury reacted with iodine, chlorine and oxygen to form a fine particulate solid. Quantitative recovery of mercury compounds formed in the vitrification system off-gas was not possible due to high (37 to 90%) deposition of solids in the off-gas piping. The behavior of mercury and iodine was most strongly influenced by the vitrification system atmosphere. During experiments performed in which the oxygen content of the vitrification system atmosphere was low (< 1 vol%); iodine retention in the glass product was 27 to 55%, the mercury composition of the solids recovered from the off-gas scrub solutions was 75 to 85 wt%, and a small quantity of metallic mercury was recovered from the off-gas scrub solution. During experiments performed in which the oxygen content of the vitrification system atmosphere was high (20 vol%), iodide retention in the glass product was 3 to 15%, the mercury composition of the solids recovered from the off-gas scrub solutions was 60 to 80 wt%, and very little metallic mercury was recovered from the off-gas scrub solution

  9. Process technology for vitrification of defense high-level waste at the Savannah River Plant

    International Nuclear Information System (INIS)

    Boersma, M.D.

    1984-01-01

    Vitrification in borosilicate glass is now the leading worldwide process for immobilizing high-level radioactive waste. Each vitrification project, however, has its unique mission and technical challenges. The Defense Waste Vitrification Facility (DWPF) now under construction at the Savannah River Plant will concentrate and vitrify a large amount of relatively low-power alkaline waste. Process research and development for the DWPF have produced significant advances in remote chemical operations, glass melting, off-gas treatment, slurry handling, decontamination, and welding. 6 references, 1 figure, 5 tables

  10. Corrosion of Metal Inclusions In Bulk Vitrification Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    Bacon, Diana H.; Pierce, Eric M.; Wellman, Dawn M.; Strachan, Denis M.; Josephson, Gary B.

    2006-07-31

    The primary purpose of the work reported here is to analyze the potential effect of the release of technetium (Tc) from metal inclusions in bulk vitrification waste packages once they are placed in the Integrated Disposal Facility (IDF). As part of the strategy for immobilizing waste from the underground tanks at Hanford, selected wastes will be immobilized using bulk vitrification. During analyses of the glass produced in engineering-scale tests, metal inclusions were found in the glass product. This report contains the results from experiments designed to quantify the corrosion rates of metal inclusions found in the glass product from AMEC Test ES-32B and simulations designed to compare the rate of Tc release from the metal inclusions to the release of Tc from glass produced with the bulk vitrification process. In the simulations, the Tc in the metal inclusions was assumed to be released congruently during metal corrosion as soluble TcO4-. The experimental results and modeling calculations show that the metal corrosion rate will, under all conceivable conditions at the IDF, be dominated by the presence of the passivating layer and corrosion products on the metal particles. As a result, the release of Tc from the metal particles at the surfaces of fractures in the glass releases at a rate similar to the Tc present as a soluble salt. The release of the remaining Tc in the metal is controlled by the dissolution of the glass matrix. To summarize, the release of 99Tc from the BV glass within precipitated Fe is directly proportional to the diameter of the Fe particles and to the amount of precipitated Fe. However, the main contribution to the Tc release from the iron particles is over the same time period as the release of the soluble Tc salt. For the base case used in this study (0.48 mass% of 0.5 mm diameter metal particles homogeneously distributed in the BV glass), the release of 99Tc from the metal is approximately the same as the release from 0.3 mass% soluble Tc

  11. In situ vitrification of transuranic wastes: systems evaluation and applications assessment

    International Nuclear Information System (INIS)

    Fitzpatrick, V.F.; Brown; Buelt, J.L.; King, S.E.; Napier, B.A.; Oma, K.H.; Silviera, D.J.; Timmerman, C.L.

    1984-01-01

    In situ vitrification is an emerging technology, suitable for the stabilization of radioactive waste. In just under three years in situ vitrification has moved from a concept tested in the laboratory to an achievable reality, with a series of 18 laboratory (engineering-scale) tests and 7 field (pilot-scale) tests. A radioactive make-up site of TRU wastes and mixed fission products has been stabilized with the pilot-scale system. In this test, approximately 25 kg of contaminated soil, containing about 600 nCi/g of 239 Pu/ 241 Am, was successfully vitrified without release of radionuclides to the environment. The ISV tests have been supplemented by public and occupational exposure calculations which show that releases fall below the limits set by the federal government for both routine and accident conditions. Performance analysis has verified the process effectiveness in terms of preventing the migration of radionuclides into the biosphere in the far-distant (1000 and 10,000 year) future under an array of intrusion scenarios including inadvertent and deliberate human intrusion. Cost analyses have shown that for both Hanford and a generic site, processing costs are less than the cost of disposal of low-level waste at U.S. Department of Energy disposal sites

  12. Hanford Waste Vitrification Plant Project advanced conceptual design summary report

    International Nuclear Information System (INIS)

    Anderson, T.D.

    1988-11-01

    The Hanford Waste Vitrification Plant (HWVP) will immobilize Hanford defense liquid high-level waste in borosilicate glass in preparation for shipment to a geologic repository. The shipment of the waste to the repository will satisfy an objective in the President's Defense Waste Management Plan. The glass product will be cast into stainless steel canisters, which will be sealed and stored at Hanford until they are shipped. This document summarizes work performed during the Advance Conceptual Design (ACD) of the HWVP. In the Reference Conceptual Design phase, which preceded the ACD, a number of design issues were identified with the potential to improve cost effectiveness, safety, constructibility, and operability. The ACD addressed and evaluated these design issues. Implementation of recommendations derived from ACD work will occur in subsequent design phases. The next design phase is preliminary design which will be followed by detailed design and construction. Net potential cost improvements of more than $36.9M were identified along with improvements in safety, constructibility, and operability. No negative schedule impacts will result from implementation of the improvements. 11 refs., 5 figs., 3 tabs

  13. ''Cold crucible'' vitrification projects for low and high active waste

    International Nuclear Information System (INIS)

    Roux, P.; Jouan, A.

    1998-01-01

    In continuity of the CEA HLW vitrification process experienced for more than 20 years in industrial operations in Cogema reprocessing plants (Marcoule and La Hague), CEA has developed an advanced extended performance cold crucible glass melter to address a wider range of waste like LLW, ILW and in particular waste with very corrosive species or requiring glass with higher elaboration temperature. In the cold crucible melter the bath of molten glass is directly heated by induction while the walls are cooled in order to freeze a protective glass layer. This technology subsequently allows high glass throughput while keeping the flexibility, the maintainability and low secondary waste generation related to a small metallic melter. Its recent use in the glass industry and the thousands of hours of pilot tests performed on inactive surrogates have demonstrated the maturity of this technology and its flexibility of use for processing most of the waste generated at nuclear facilities. SGN has therefore proposed this technology in Italy and Korea and in USA in the frame of the Hanford Privatization phase 1 A feasibility study. Main features of this study but also tests results with Hanford surrogates and active samples are discussed. (author)

  14. STATUS and DIRECTION OF THE BULK VITRIFICATION PROGRAM FOR THE SUPPLEMENTAL TREATMENT OF LOW ACTIVITY TANK WASTE AT HANFORD

    International Nuclear Information System (INIS)

    RAYMOND, R.E.

    2005-01-01

    The DOE Office of River Protection (ORP) is managing a program at the Hanford site that will retrieve and treat more than 200 million liters (53 million gal.) of radioactive waste stored in underground storage tanks. The waste was generated over the past 50 years as part of the nation's defense programs. The project baseline calls for the waste to be retrieved from the tanks and partitioned to separate the highly radioactive constituents from the large volumes of chemical waste. These highly radioactive components will be vitrified into glass logs in the Waste Treatment Plant (WTP), temporarily stored on the Hanford Site, and ultimately disposed of as high-level waste in the offsite national repository. The less radioactive chemical waste, referred to as low-activity waste (LAW), is also planned to be vitrified by the WTP, and then disposed of in approved onsite trenches. However, additional treatment capacity is required in order to complete the pretreatment and immobilization of the tank waste by 2028, which represents a Tri-Party Agreement milestone. To help ensure that the treatment milestones will be met, the Supplemental Treatment Program was undertaken. The program, managed by CH2M HILL Hanford Group, Inc., involves several sub-projects each intended to supplement part of the treatment of waste being designed into the WTP. This includes the testing, evaluation, design, and deployment of supplemental LAW treatment and immobilization technologies, retrieval and treatment of mixed TRU waste stored in the Hanford Tanks, and supplemental pre-treatment. Applying one or more supplemental treatment technologies to the LAW has several advantages, including providing additional processing capacity, reducing the planned loading on the WTP, and reducing the need for double-shell tank space for interim storage of LAW. In fiscal year 2003, three potential supplemental treatment technologies were evaluated including grout, steam reforming and bulk vitrification using AMEC

  15. Off-Gas Analysis During the Vitrification of Hanford Radioactive Waste Samples

    International Nuclear Information System (INIS)

    Ha, B.C.; Ferrara, D.M.; Crawford, C.L.; Choi, A.S.; Bibler, N.E.

    1998-01-01

    This paper describes the off-gas analysis of samples collected during the radioactive vitrification experiments. Production and characterization of the Hanford waste-containing LAW and HAW glasses are presented in related reports from this conference

  16. Hanford Waste Vitrification Plant quality assurance program description: Overview and applications

    International Nuclear Information System (INIS)

    Caplinger, W.H.

    1990-12-01

    This document describes the Hanford Waste Vitrification Plant Project Quality Assurance Program. This program is being implemented to ensure the acceptability of high-level radioactive canistered waste forms produced by the Hanford Waste Vitrification Plant for disposal in a licensed federal repository. The Hanford Waste Vitrification Plant Quality Assurance Program is comprised of this Quality Assurance Program Description as well as the associated contractors' quality assurance programs. The objective of this Quality Assurance Program Description is to provide the Hanford Waste Vitrification Plant Project participants with guidance and direction for program implementation while satisfying the US Department of Energy Office of Civilian Radioactive Waste Management needs in repository licensing activities with regard to canistered waste forms. To accomplish this objective, this description will be prepared in three parts: Part 1 - Overview and applications document; Part 2 - Development and qualification of the canistered waste form; Part 3 - Production of canistered waste forms. Part 1 describes the background, strategy, application, and content of the Hanford Waste Vitrification Plant Quality Assurance Program. This Quality Assurance Program Description, when complete, is designed to provide a level of confidence in the integrity of the canistered waste forms. 8 refs

  17. A rapid and continuous system for immobilization of nuclear material waste by vitrification

    International Nuclear Information System (INIS)

    Shareef, M.U.; Hussain, S.H.; Tufail, M.; Rashid, F.

    2009-01-01

    The nuclear technology has prime importance and backbone for rapid development of medical sciences, industries and in power generation as an alternate source of energy. Despite all these facts there is a major problem which is always associated with nuclear technology, that is, the generation of undesirable radioactive wastes. The radioactive wastes are quite problematic and need major attention for its treatment, conditioning and properly disposal to keep the environmental activities and human ecosystem healthy and safe. There are different large scale methods and processes to treat and dispose off the radioactive wastes. These processes are evaluated and designed by the various world competent and pronounced scientists in the light of rules and safety limits set by IAEA and other regulatory authorities to protect the environment and eventually protect our ecosystem. The research and development work on radioactive waste has been proceeding for the last fifty years but still it is a core issue and a big challenge for the nuclear scientists and radiation workers. In this study a rapid and continuous system for immobilization of nuclear waste into glass matrix by vitrification has been designed. In general treatment methods, Borosilicate glass is preferred because it is efficient, cost effective and rapid to that of other radioactive waste form. In this process the simulated waste is mixed with glass forming material and process for melting to form a glassy substrate in continuous manner. The waste is being converted into vitreous form and encapsulate into a glass matrix. (author)

  18. Vitrification testing of simulated high-level radioactive waste at Hanford

    International Nuclear Information System (INIS)

    Perez, J.M. Jr.; Nakaoka, R.R.

    1986-03-01

    The Hanford Waste Vitrification Plant may apply vitrification technology, being developed at Pacific Northwest Laboratory, to solidify selected Hanford waste streams prior to disposal in a federal repository. Based on the first stage of flowsheet development and laboratory testing, a reference working glass and two candidate simulated feed slurries were recommended for vitrification testing. Over 500 hours of melter testing were performed in 1985 during prototype vitrification experiments. Testing demonstrated that the slurry compositions had acceptable processing characteristics in a ceramic melter. A pre-made glass-former frit was determined to be preferred as the method of glass-former addition. Due to a high chromium content in the waste, spinal crystal formation and settling occurred in the glass tank. The nature and extent of off-gas effluents were consistent with past experiments processing slurries containing formic acid

  19. Nuclear Waste Vitrification in the U.S.: Recent Developments and Future Options

    International Nuclear Information System (INIS)

    Vienna, John D.

    2010-01-01

    Nuclear power plays a key role in maintaining current world wide energy growth while minimizing the greenhouse gas emissions. A disposition path for used nuclear fuel (UNF) must be found for this technology to achieve its promise. One likely option is the recycling of UNF and immobilization of the high-level waste (HLW) by vitrification. Vitrification is the technology of choice for immobilizing HLW from defense and commercial fuel reprocessing around the world. Recent advances in both recycling technology and vitrification show great promise in closing the nuclear fuel cycle in an efficient and economical fashion. This article summarizes the recent trends developments and future options in waste vitrification for both defense waste cleanup and closing the nuclear fuel cycle in the U.S.

  20. Mixed waste management options

    International Nuclear Information System (INIS)

    Owens, C.B.; Kirner, N.P.

    1992-01-01

    Currently, limited storage and treatment capacity exists for commercial mixed waste streams. No commercial mixed waste disposal is available, and it has been estimated that if and when commercial mixed waste disposal becomes available, the costs will be high. If high disposal fees are imposed, generators may be willing to apply extraordinary treatment or regulatory approaches to properly dispose of their mixed waste. This paper explores the feasibility of several waste management scenarios and management options. Existing data on commercially generated mixed waste streams are used to identify the realm of mixed waste known to be generated. Each waste stream is evaluated from both a regulatory and technical perspective in order to convert the waste into a strictly low-level radioactive or a hazardous waste. Alternative regulatory approaches evaluated in this paper include a delisting petition) no migration petition) and a treatability variance. For each waste stream, potentially available treatment options are identified that could lead to these variances. Waste minimization methodology and storage for decay are also considered. Economic feasibility of each option is discussed broadly. Another option for mixed waste management that is being explored is the feasibility of Department of Energy (DOE) accepting commercial mixed waste for treatment, storage, and disposal. A study has been completed that analyzes DOE treatment capacity in comparison with commercial mixed waste streams. (author)

  1. Hanford Waste Vitrification Plant quality assurance program description for defense high-level waste form development and qualification

    International Nuclear Information System (INIS)

    Hand, R.L.

    1990-12-01

    The US Department of Energy-Office of Civilian Radioactive Waste Management has been designated the national high-level waste repository licensee and the recipient for the canistered waste forms. The Office of Waste Operations executes overall responsibility for producing the canistered waste form. The Hanford Waste Vitrification Plant Project, as part of the waste form producer organization, consists of a vertical relationship. Overall control is provided by the US Department of Energy-Environmental Restoration and Waste Management Headquarters; with the US Department of Energy-Office of Waste Operations; the US Department of Energy- Headquarters/Vitrification Project Branch; the US Department of Energy-Richland Operations Office/Vitrification Project Office; and the Westinghouse Hanford Company, operations and engineering contractor. This document has been prepared in response to direction from the US Department of Energy-Office of Civilian Radioactive Waste Management through the US Department of Energy-Richland Operations Office for a quality assurance program that meets the requirements of the US Department of Energy. This document provides guidance and direction for implementing a quality assurance program that applies to the Hanford Waste Vitrification Plant Project. The Hanford Waste Vitrification Plant Project management commits to implementing the quality assurance program activities; reviewing the program periodically, and revising it as necessary to keep it current and effective. 12 refs., 6 figs., 1 tab

  2. MIxed Waste Integrated Program (MWIP): Technology summary

    International Nuclear Information System (INIS)

    1994-02-01

    The mission of the Mixed Waste Integrated Program (MWIP) is to develop and demonstrate innovative and emerging technologies for the treatment and management of DOE's mixed low-level wastes (MLLW) for use by its customers, the Office of Waste Operations (EM-30) and the Office of Environmental Restoration (EM-40). The primary goal of MWIP is to develop and demonstrate the treatment and disposal of actual mixed waste (MMLW and MTRU). The vitrification process and the plasma hearth process are scheduled for demonstration on actual radioactive waste in FY95 and FY96, respectively. This will be accomplished by sequential studies of lab-scale non-radioactive testing followed by bench-scale radioactive testing, followed by field-scale radioactive testing. Both processes create a highly durable final waste form that passes leachability requirements while destroying organics. Material handling technology, and off-gas requirements and capabilities for the plasma hearth process and the vitrification process will be established in parallel

  3. Temperature Distribution within a Cold Cap during Nuclear Waste Vitrification.

    Science.gov (United States)

    Dixon, Derek R; Schweiger, Michael J; Riley, Brian J; Pokorny, Richard; Hrma, Pavel

    2015-07-21

    The kinetics of the feed-to-glass conversion affects the waste vitrification rate in an electric glass melter. The primary area of interest in this conversion process is the cold cap, a layer of reacting feed on top of the molten glass. The work presented here provides an experimental determination of the temperature distribution within the cold cap. Because direct measurement of the temperature field within the cold cap is impracticable, an indirect method was developed in which the textural features in a laboratory-made cold cap with a simulated high-level waste feed were mapped as a function of position using optical microscopy, scanning electron microscopy, energy dispersive spectroscopy, and X-ray diffraction. The temperature distribution within the cold cap was established by correlating microstructures of cold-cap regions with heat-treated feed samples of nearly identical structures at known temperatures. This temperature profile was compared with a mathematically simulated profile generated by a cold-cap model that has been developed to assess the rate of glass production in a melter.

  4. Removal of dissolved and suspended radionuclides from Hanford Waste Vitrification Plant liquid wastes

    International Nuclear Information System (INIS)

    Sharp, S.D.; Nankani, F.D.; Bray, L.A.; Eakin, D.E.; Larson, D.E.

    1990-12-01

    It was determined during Preliminary Design of the Hanford Waste Vitrification Plant that certain intermediate process liquid waste streams should be decontaminated in a way that would permit the purge of dissolved chemical species from the process recycle shop. This capability is needed to ensure proper control of product glass chemical composition and to avoid excessive corrosion of process equipment. This paper discusses the process design of a system that will remove both radioactive particulates and certain dissolved fission products from process liquid waste streams. Supporting data obtained from literature sources as well as from laboratory- and pilot-scale tests are presented. 3 refs., 1 fig., 3 tabs

  5. Glass optimization for vitrification of Hanford Site low-level tank waste

    International Nuclear Information System (INIS)

    Feng, X.; Hrma, P.R.; Westsik, J.H. Jr.

    1996-03-01

    The radioactive defense wastes stored in 177 underground single-shell tanks (SST) and double-shell tanks (DST) at the Hanford Site will be separated into low-level and high-level fractions. One technology activity underway at PNNL is the development of glass formulations for the immobilization of the low-level tank wastes. A glass formulation strategy has been developed that describes development approaches to optimize glass compositions prior to the projected LLW vitrification facility start-up in 2005. Implementation of this strategy requires testing of glass formulations spanning a number of waste loadings, compositions, and additives over the range of expected waste compositions. The resulting glasses will then be characterized and compared to processing and performance specifications yet to be developed. This report documents the glass formulation work conducted at PNL in fiscal years 1994 and 1995 including glass formulation optimization, minor component impacts evaluation, Phase 1 and Phase 2 melter vendor glass development, liquidus temperature and crystallization kinetics determination. This report also summarizes relevant work at PNNL on high-iron glasses for Hanford tank wastes conducted through the Mixed Waste Integrated Program and work at Savannah River Technology Center to optimize glass formulations using a Plackett-Burnam experimental design

  6. Status of the French AVM vitrification facility

    International Nuclear Information System (INIS)

    Bonniaud, R.A.; Jouan, A.F.; Sombret, C.G.

    1979-01-01

    The Commission of the Marcoule Vitrification Plant (or AVM) has opened the industrial development era for the continuous vitrification process. Radioactive liquid wastes are calcinated in a rotary kiln to give a solid form, mixed with suitable raw materials in an electric furnace to make the glass. The glass is poured in containers and transferred to a disposal facility. The off gas released are processed. Design of La Hague next vitrification plant is given

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

    International Nuclear Information System (INIS)

    Howden, G.F.

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Howden, G.F.

    1994-10-24

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

  9. Remote maintenance demonstration tests at a pilot plant for high level waste vitrification

    International Nuclear Information System (INIS)

    Selig, M.

    1984-01-01

    The remote maintenance and replacement technique designed for a radioactive vitrification plant have been developed and tested in a full scale handling mockup and in an inactive pilot plants by the Central Engineering Department of the Karlsruhe Nuclear Research Center. As a result of the development work and the tests it has been proved that the remote maintenance technique and remote handling equipment can be used without any technical problems and are suited for application in a radioactive waste vitrification plant

  10. Solidification and vitrification life-cycle economics study

    International Nuclear Information System (INIS)

    Gimpel, R.F.

    1992-01-01

    Solidification (making concrete) and vitrification (making glass) are frequently the treatment methods recommended for treating inorganic or radioactive wastes. Ex-situ solidification and vitrification are the competing methods for treating in excess of 450 000 cm 3 of low-level radioactive and mixed wastes at the Fernald Environmental Management Project (FEMP) located near Cincinnati, Ohio. This paper summarizes a detailed study done to: (1) compare the economics of the solidification and vitrification processes, (2) determine if the stigma assigned to vitrification is warranted and, (3) determine if investing millions of dollars into vitrification development, along with solidification development, at Fernald is warranted

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

    International Nuclear Information System (INIS)

    SCHAUS, P.S.

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    SCHAUS, P.S.

    2006-07-21

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

  13. An update on the quality assurance for the waste vitrification plants

    Energy Technology Data Exchange (ETDEWEB)

    Caplinger, W.H.; Shugars, D.L.; Carlson, M.K.

    1990-01-01

    Immobilization of high-level defense production wastes is an important step in environmental restoration. The best available technology for immobilization of this waste currently is by incorporation into borosilicate glass, i.e., vitrification. Three US sites are active in the design, construction, or operation of vitrification facilities. The status, facility description and Quality Assurance (QA) development for each facility was presented at the 1989 Energy Division Conference. This paper presents the developments since that time. The West Valley Demonstration Project (WVDP) in northwestern New York State has demonstrated the technology. At the Savannah River Site (SRS) in South Carolina the Defense Waste Processing Facility (DWPF) has completed design, construction is essentially complete, and preparation for operation is underway. The Hanford Waste Vitrification Plant (HWVP) in Washington State is in initial Detailed Design. 4 refs.

  14. An update on the quality assurance for the waste vitrification plants

    International Nuclear Information System (INIS)

    Caplinger, W.H.; Shugars, D.L.; Carlson, M.K.

    1990-01-01

    Immobilization of high-level defense production wastes is an important step in environmental restoration. The best available technology for immobilization of this waste currently is by incorporation into borosilicate glass, i.e., vitrification. Three US sites are active in the design, construction, or operation of vitrification facilities. The status, facility description and Quality Assurance (QA) development for each facility was presented at the 1989 Energy Division Conference. This paper presents the developments since that time. The West Valley Demonstration Project (WVDP) in northwestern New York State has demonstrated the technology. At the Savannah River Site (SRS) in South Carolina the Defense Waste Processing Facility (DWPF) has completed design, construction is essentially complete, and preparation for operation is underway. The Hanford Waste Vitrification Plant (HWVP) in Washington State is in initial Detailed Design. 4 refs

  15. Interaction analysis method for the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Grant, P.R.; Deshotels, R.L.; Van Katwijk, C.

    1993-01-01

    In order to anticipate potential problems as early as possible during the design effort, a method for interaction analysis was developed to meet the specific hazards of the Hanford Waste Vitrification Plant (HWVP). The requirement for interaction analysis is given in DOE Order 6430.1B and DOE-STD-1021-92. The purpose of the interaction analysis is to ensure that non-safety class items will not fail in a manner that will adversely affect the ability of any safety class item to perform its safety function. In the HWVP there are few structures, equipment, or controls that are safety class (those with a direct safety function, i.e., confinement of waste). In addition to damage due to failure of non-safety class items as a result of natural phenomena, threats to HWVP safety class items include the following: room flooding from firewater, leakage of chemically reactive liquids, high-pressure gas impingement from leaking piping, rocket-type impact from broken pressurized gas cylinders, loss of control of mobile equipment, cryogenic liquid spill, fire, and smoke. The time needed to perform the interaction analysis is minimized by consolidating safety class items into segregated areas. Each area containing safety class items is evaluated, and any potential threat to the safety functions is noted. After relocation of safety class items is considered, items that pose a threat are generally upgraded to eliminate the threat to the safety class items. Upgraded items are designed to not fail under the conditions being evaluated. Upgrading is the preferred option when relocation is not possible. Other options are to provide barriers, design the safety class item not to be damaged by failed items, or rely on redundancy and isolation from local threats. The upgraded features of non-safety class items are designed to the same quality standards as the safety class items

  16. Vitrification technology for treating low-level waste from nuclear facilities

    International Nuclear Information System (INIS)

    Oniki, Toshiro; Nabemoto, Toyonobu; Fukui, Toshiki

    2016-01-01

    The development of technologies for treating nuclear waste generated by nuclear power plants and reprocessing plants during their operation or decommissioning is underway both in Japan and abroad. Of the many types of treatment technologies that have been developed, vitrification technology is attracting attention as being the most promising technology for converting such waste into a stable state. As a brief review of technical developments aimed at reducing nuclear waste and finding a solution to the final disposal issue, this paper describes approaches to completing the development of vitrification technology in Japan, including IHI's activities. (author)

  17. Mixed Waste Management Facility

    International Nuclear Information System (INIS)

    Brummond, W.; Celeste, J.; Steenhoven, J.

    1993-08-01

    The DOE has developed a National Mixed Waste Strategic Plan which calls for the construction of 2 to 9 mixed waste treatment centers in the Complex in the near future. LLNL is working to establish an integrated mixed waste technology development and demonstration system facility, the Mixed Waste Management Facility (MWMF), to support the DOE National Mixed Waste Strategic Plan. The MWMF will develop, demonstrate, test, and evaluate incinerator-alternatives which will comply with regulations governing the treatment and disposal of organic mixed wastes. LLNL will provide the DOE with engineering data for design and operation of new technologies which can be implemented in their mixed waste treatment centers. MWMF will operate under real production plant conditions and process samples of real LLNL mixed waste. In addition to the destruction of organic mixed wastes, the development and demonstration will include waste feed preparation, material transport systems, aqueous treatment, off-gas treatment, and final forms, thus making it an integrated ''cradle to grave'' demonstration. Technologies from offsite as well as LLNL's will be tested and evaluated when they are ready for a pilot scale demonstration, according to the needs of the DOE

  18. Bulk Vitrification Technology For The Treatment And Immobilization Of Low-Activity Waste

    International Nuclear Information System (INIS)

    Ard, K.E.

    2011-01-01

    This report is one of four reports written to provide background information regarding immobilization technologies under consideration for supplemental immobilization of Hanford's low-activity waste. This paper is intended to provide the reader with general understanding of Bulk Vitrification and how it might be applied to immobilization of Hanford's low-activity waste.

  19. BULK VITRIFICATION TECHNOLOGY FOR THE TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE

    Energy Technology Data Exchange (ETDEWEB)

    ARD KE

    2011-04-11

    This report is one of four reports written to provide background information regarding immobilization technologies under consideration for supplemental immobilization of Hanford's low-activity waste. This paper is intended to provide the reader with general understanding of Bulk Vitrification and how it might be applied to immobilization of Hanford's low-activity waste.

  20. High-level waste vitrification: the state of the art in France

    International Nuclear Information System (INIS)

    Sombret, C.; Maillet, J.

    1988-02-01

    This paper describes the main features of the French high-level waste vitrification program. These features include: - extensive R and D for more than 20 years; - successful operation of the AVM facility at Marcoule for about 10 years; - startup of six vitrification lines at La Hague, in the near future. The CEA is pursuing R and D for mid-term vitrification enhancement. New R and D facilities are being built at Marcoule to increase the capacity of vitrification equipment, study glass preparation at even higher temperatures to increase SiO 2 and Al 2 O 3 concentration, and perform extensive testing of samples with very high activity (more than 5,000Ci/l). 8 refs

  1. Utilization of the Pilot Scale Demonstration Facility for Vitrification of Low and Intermediate Level Radioactive Wastes

    International Nuclear Information System (INIS)

    Oh, Won Zin; Choi, W. K.; Jung, C. H.; Won, H. J.; Song, P. S.; Min, B. Y.; Park, H. S.; Jung, K. K.; Yun, K. S.

    2005-10-01

    A series of maintenance and repair work for normalization of the pilot scale vitrification demonstration facility was completed successfully to develop the waste treatment in high temperature and melting technology. It was investigated that the treatment of combustible and non-combustible wastes produced at the KAERI site is technically feasible in the pilot scale vitrification demonstration facility which is designed to be able to treat various kinds of radioactive wastes such as combustible and non-combustible wastes including soil and concrete. The vitrification test facility can be used as the R and D and the technology demonstration facility for melt decontamination of the metallic wastes which have a fixed specification. The modification of the RI storage room in the pilot scale vitrification demonstration facility and the licensing according to the facility modification were completed for the R and D on melt decontamination of dismantled metallic wastes which is carrying out as one of the national long-term R and D projects on nuclear energy. The lab-scale melt decontamination apparatus was installed in modified RI storage room and the characteristics of melt decontamination will be examined using various metallic wastes. It is expected that the economical feasibility on the volume reduction and recycle of metallic wastes will be escalated in the present situation when the unit cost for waste disposal has the tendency to grow up gradually. Therefore, the pilot scale vitrification demonstration facility can be used for the technology development for the volume reduction and recycle of the metallic wastes generated from on-going projects on the decommissioning of research reactors and the environmental restoration of uranium conversion plant, and for the treatment of radioactive solid wastes produced at the KAERI site

  2. Utilization of the Pilot Scale Demonstration Facility for Vitrification of Low and Intermediate Level Radioactive Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Won Zin; Choi, W. K.; Jung, C. H.; Won, H. J.; Song, P. S.; Min, B. Y.; Park, H. S.; Jung, K. K.; Yun, K. S

    2005-10-15

    A series of maintenance and repair work for normalization of the pilot scale vitrification demonstration facility was completed successfully to develop the waste treatment in high temperature and melting technology. It was investigated that the treatment of combustible and non-combustible wastes produced at the KAERI site is technically feasible in the pilot scale vitrification demonstration facility which is designed to be able to treat various kinds of radioactive wastes such as combustible and non-combustible wastes including soil and concrete. The vitrification test facility can be used as the R and D and the technology demonstration facility for melt decontamination of the metallic wastes which have a fixed specification. The modification of the RI storage room in the pilot scale vitrification demonstration facility and the licensing according to the facility modification were completed for the R and D on melt decontamination of dismantled metallic wastes which is carrying out as one of the national long-term R and D projects on nuclear energy. The lab-scale melt decontamination apparatus was installed in modified RI storage room and the characteristics of melt decontamination will be examined using various metallic wastes. It is expected that the economical feasibility on the volume reduction and recycle of metallic wastes will be escalated in the present situation when the unit cost for waste disposal has the tendency to grow up gradually. Therefore, the pilot scale vitrification demonstration facility can be used for the technology development for the volume reduction and recycle of the metallic wastes generated from on-going projects on the decommissioning of research reactors and the environmental restoration of uranium conversion plant, and for the treatment of radioactive solid wastes produced at the KAERI site.

  3. DOE regulatory reform initiative vitrified mixed waste

    International Nuclear Information System (INIS)

    Carroll, S.J.; Holtzscheiter, E.W.

    1997-01-01

    The US Department of Energy (DOE) is charged with responsibly managing the largest volume of mixed waste in the United States. This responsibility includes managing waste in compliance with all applicable Federal and State laws and regulations, and in a cost-effective, environmentally responsible manner. Managing certain treated mixed wastes in Resource Conservation and Recovery Act (RCRA) permitted storage and disposal units (specifically those mixed wastes that pose low risks from the hazardous component) is unlikely to provide additional protection to human health and the environment beyond that afforded by managing these wastes in storage and disposal units subject to requirements for radiological control. In October, 1995, the DOE submitted a regulatory reform proposal to the Environmental Protection Agency (EPA) relating to vitrified mixed waste forms. The technical proposal supports a regulatory strategy that would allow vitrified mixed waste forms treated through a permit or other environmental compliance mechanism to be granted an exemption from RCRA hazardous waste regulation, after treatment, based upon the inherent destruction and immobilization capabilities of vitrification technology. The vitrified waste form will meet, or exceed the performance criteria of the Environmental Assessment (EA) glass that has been accepted as an international standard for immobilizing radioactive waste components and the LDR treatment standards for inorganics and metals for controlling hazardous constituents. The proposal further provides that vitrified mixed waste would be responsibly managed under the Atomic Energy Act (AEA) while reducing overall costs. Full regulatory authority by the EPA or a State would be maintained until an acceptable vitrified mixed waste form, protective of human health and the environment, is produced

  4. Mechanism of ruthenium dioxide crystallization during high level waste vitrification

    International Nuclear Information System (INIS)

    Boucetta, H.

    2012-01-01

    Ruthenium, arising from the reprocessing of spent uranium oxide fuel, has a low solubility in glass melt. It crystallizes in the form of particles of RuO 2 of acicular or polyhedral morphology dispersed in fission product and actinides waste containment glass. Since the morphology of these particles strongly influences the physico-chemical properties, the knowledge and the control of their mechanism of formation are of major importance. The goal of this work is to determine the chemical reactions responsible for the formation of RuO 2 particles of acicular or polyhedral shape during glass synthesis. Using a simplification approach, the reactions between RuO 2 -NaNO 3 , and more complex calcine RuO 2 -Al 2 O 3 -Na 2 O and a sodium borosilicate glass are studied. In situ scanning electron microscopy and XANES at increasing temperatures are used to follow changes in composition, speciation and morphology of the ruthenium intermediate species. Those compounds are thoroughly characterised by SEM, XRD, HRTEM, and ruthenium K-edge X-ray absorption spectroscopy. This combined approach allows us to show that the ruthenium speciation modification during vitrification is the key of control of the morphology of RuO 2 particles in the glass. In particular, the formation of a specific intermediate compound (Na 3 RuO 4 ) is one of the main steps that lead to the precipitation of needle-shaped RuO 2 particles in the melt. The formation of polyhedral particles, on the contrary, results from the direct incorporation of RuO 2 crystals in the melt followed by an Ostwald ripening mechanism. (author) [fr

  5. Pecularities of carrying out radioactive wastes vitrification process without preliminary calcination of wastes

    International Nuclear Information System (INIS)

    Konstantinovich, A.A.; Kulichenko, V.V.; Bel'tyukov, V.A.; Nikiforov, A.S.; Nikipelov, B.V.; Stepanov, S.E.; Baskov, L.I.; Kulakov, S.I.

    1978-01-01

    Vitrification technology is considered for liquid radioactive wastes by means of electric furnace where heating of glass-paste is done by electric current passing through the melt. Continious process of gehydration, calcination and vitrification is going on in one apparatus. Testing if the method has been performed by use of a model solution, containing sodium and aluminium nitrates. To obtain phosphoric acid has been added into the solution. Lay-out of the device and its description as well as technical parameters of the electric furnace are given. The results are stated for determination of the optimum operation conditions for the device. To reduce entrainment of solid components, molasses has been added in the solution. Parameters are given for the process of the solution containing 80 g/l molasses processing. It has been shown that edding molasses to the solution permitted to reduse power consumption of the process due to the heat generation during oxidation-reduction reaction on the melt surface. The results are given for investigations of the nitrogen oxides catching in scrubbers. These results have shown that introduction of molasses reduces nitrigen oxides concentration. The results of the experimental works have shown the possibility of the continious process of dehydration, calcination and vitrification in single device with application of remote control and monitoring by means of automatics. (I.T.) [ru

  6. Melter system technology testing for Hanford Site low-level tank waste vitrification

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1996-01-01

    Following revisions to the Tri-Party Agreement for Hanford Site cleanup, which specified vitrification for Complete melter feasibility and system operability immobilization of the low-level waste (LLW) tests, select reference melter(s), and establish reference derived from retrieval and pretreatment of the radioactive LLW glass formulation that meets complete systems defense wastes stored in 177 underground tanks, commercial requirements (June 1996). Available melter technologies were tested during 1994 to 1995 as part of a multiphase program to select reference Submit conceptual design and initiate definitive design technologies for the new LLW vitrification mission

  7. Mercury reduction and removal during high-level radioactive waste processing and vitrification

    International Nuclear Information System (INIS)

    Eibling, R.E.; Fowler, J.R.

    1981-01-01

    A reference process for immobilizing the high-level radioactive waste in borosilicate glass has been developed at the Savannah River Plant. This waste contains a substantial amount of mercury from separations processing. Because mercury will not remain in borosilicate glass at the processing temperature, mercury must be removed before vitrification or must be handled in the off-gas system. A process has been developed to remove mercury by reduction with formic acid prior to vitrification. Additional benefits of formic acid treatment include improved sludge handling and glass melter redox control

  8. Startup and operation of a plant-scale continuous glass melter for vitrification of Savannah River Plant simulated waste

    International Nuclear Information System (INIS)

    Willis, T.A.

    1980-01-01

    The reference process for disposal of radioactive waste from the Savannah River Plant is vitrification of the waste in borosilicate glass in a continuous glass melter. Design, startup, and operation of a plant-scale developmental melter system are discussed

  9. Guidelines for mixed waste minimization

    International Nuclear Information System (INIS)

    Owens, C.

    1992-02-01

    Currently, there is no commercial mixed waste disposal available in the United States. Storage and treatment for commercial mixed waste is limited. Host States and compacts region officials are encouraging their mixed waste generators to minimize their mixed wastes because of management limitations. This document provides a guide to mixed waste minimization

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

    International Nuclear Information System (INIS)

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

    1992-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    GR Golcar; NG Colton; JG Darab; HD Smith

    2000-04-04

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

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

    International Nuclear Information System (INIS)

    GR Golcar; NG Colton; JG Darab; HD Smith

    2000-01-01

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

  13. Installation and routing of critical embedments at the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Van Katwijk, C.; Keenan, R.M.; Watts, C.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) is being designed by Fluor Daniel. Waste Chem Corporation is providing specialized expertise as Fluor Daniel's major subcontractor for vitrification and remote systems technologies. Westinghouse Hanford Company (Westinghouse Hanford) is the Project Integration manager and Business manager, and as the plant operator it provides technical direction to the Architect/Engineer team and constructor on behalf of the US Department of Energy, Richland Field Office. The Hot Cell portion of HWVP Vitrification Building contains very congested piping systems in the walls that penetrate in to the cells to nozzles for remote piping jumper assemblies. These nozzles require very tight tolerances to ensure a leak-tight fit to the jumpers. An approach has been developed that minimizes the time and expense of installing these nozzles in the wall to tight construction tolerances. This approach is called the Ganged Embed Plate (GEP) design

  14. Radioactive mixed waste disposal

    International Nuclear Information System (INIS)

    Jasen, W.G.; Erpenbeck, E.G.

    1993-02-01

    Various types of waste have been generated during the 50-year history of the Hanford Site. Regulatory changes in the last 20 years have provided the emphasis for better management of these wastes. Interpretations of the Atomic Energy Act of 1954 (AEA), the Resource Conservation and Recovery Act of 1976 (RCRA), and the Hazardous and Solid Waste Amendments (HSWA) have led to the definition of radioactive mixed wastes (RMW). The radioactive and hazardous properties of these wastes have resulted in the initiation of special projects for the management of these wastes. Other solid wastes at the Hanford Site include low-level wastes, transuranic (TRU), and nonradioactive hazardous wastes. This paper describes a system for the treatment, storage, and disposal (TSD) of solid radioactive waste

  15. Laboratory characterization and vitrification of Hanford radioactive high-level waste

    International Nuclear Information System (INIS)

    Tingey, J.M.; Elliott, M.L.; Larson, D.E.; Morrey, E.V.

    1991-05-01

    Radioactive high-level wastes generated at the Department of Energy's Hanford Site are stored in underground carbon steel tanks. Two double-shell tanks contain neutralized current acid waste (NCAW) from the reprocessing of irradiated nuclear fuel in the Plutonium and Uranium Extraction (PUREX) Plant. The tanks were sampled for characterization and waste immobilization process/product development. The high-level waste generated in PUREX was denitrated with sugar to form current acid waste (CAW). The CAW was ''neutralized'' to a pH of approximately 14 by adding sodium hydroxide to reduce corrosion of the tanks. This ''neutralized'' waste is called Neutralized Current Acid Waste. Both precipitated solids and liquids are stored in the NCAW waste tanks. The NCAW contains small amounts of plutonium and most of the fission products and americium from the irradiated fuel. NCAW also contains stainless steel corrosion products, and iron and sulfate from the ferrous sulfamate reductant used in the PUREX process. The NCAW will be retrieved, pretreated, and immobilized prior to final disposal. Pretreatment consists of water washing the precipitated NCAW solids for sulfate and soluble salts removal as a waste reduction step prior to vitrification. This waste is expected to be the first waste type to be retrieved and vitrified in the Hanford Waste Vitrification Plant (HWVP). A characterization plan was developed that details the processing of the small-volume NCAW samples through retrieval, pretreatment and vitrification process steps. Physical, rheological, chemical, and radiochemical properties were measured throughout these process steps. The results of nonradioactive simulant tests were used to develop appropriate pretreatment and vitrification process steps. The processing and characterization of simulants and actual NCAW tank samples are used to evaluate the operation of these processes. 3 refs., 1 fig., 4 tabs

  16. An experimental feasibility study on vitrification of Low - and medium-level radioactive waste

    International Nuclear Information System (INIS)

    Park, Jongkil; Song, Myungjae; Choe, Youngson; Cho, Myungyul

    1996-01-01

    Laboratory and pilot tests(all cold tests) were carried out to examine the possibility of vitrification of low-level radioactive waste such as combustible DAW(protection clothes and vinyl seat), ion exchange resins, and evaporator bottoms with three types of vitrification equipment. Pyrolyzed or dried waste material and glass formers were fed into the melting cavity, converted to molten glassy mixture, and poured into a canister. For examination of the optimal ash contents in borosilicate glass waste forms with respect to waste types, compressive strength tests were conducted for several samples of ash contents. In the case of protection clothes, vinyl seat, and spent resin was rapidly reduced up to 5 or 6 times lower than that of neat glass, but hardly changed for dried evaporator bottoms. In order to investigate the possibility of direct vitrification, combustible DAW and spent resin were directly fed into the in-can melter and Pt crucible. Pilot scale joule-heated melter in which plate type electrodes were employed to generate heat and whose melting cavity maintained a near constant molten glass level throughout the vitrification process, was designed and constructed. The total amount of molten glass in the melter was about 125 Kg and the average processing rate was 10 ∼ 15 Kg/h. At least 10 hr of retention time was considered for the best quality of the glassy waste form throughout the long-term tests

  17. In situ vitrification - A potential remedial action technique for hazardous wastes

    International Nuclear Information System (INIS)

    Fitzpatrick, V.F.; Buelt, J.L.; Oma, K.H.; Timmerman, C.L.

    1984-01-01

    In situ vitrification (ISV) is an innovative technology being developed as a potential method for stabilizing transuranic (TRU) contaminated wastes in place. Although the process is being developed for TRU contaminated wastes, it is envisioned that the process could also be applied to hazardous chemical wastes. In situ vitrification (ISV) is the conversion of contaminated soil into a durable glass and crystalline wastes form through melting by joule heating. The technology for in situ vitrification is based upon electric melter technology developed at the Pacific Northwest Laboratory (PNL) for the immobilization of high-level nuclear waste. In situ vitrification was initially tested by researchers at PNL in August, 1980 (U.S. Patent 4,376,598). Since then, ISV has grown from a concept to an emerging technology through a series of 21 engineering-scale (laboratory) tests and 7 pilot-scale (field) tests. A large-scale system is currently being fabricated for testing. The program has been sponsored by the U.S. Department of Energy's (DOE) Richland Operations Office for potential application to Hanford TRU contaminated soil sites. A more detailed description outlining the power system design and the off-gas treatment system follows

  18. Task 20 - Prevention of Chloride Corrosion in High-Temperature Waste Treatment Systems (Corrosives Removals from Vitrification Slurries)

    International Nuclear Information System (INIS)

    Timpe, R.C.; Aulich, T.R.

    1998-01-01

    GTS Duratek is working with BNFL Incorporated on a US Department of Energy (DOE) contract to develop a facility to treat and immobilize radioactive waste at the Hanford site in southeast Washington. Development of the 10-ton/day Hanford facility will be based on findings from work at Duratek's 3.3-ton/day pilot plant in Columbia, Maryland, which is in the final stage of construction and scheduled for shakedown testing in early 1999. In prior work with the Catholic University of America Vitreous State Laboratory, Duratek has found that slurrying is the most efficient way to introduce low-level radioactive, hazardous, and mixed wastes into vitrification melters. However, many of the Hanford tank wastes to be vitrified contain species (primarily chloride and sulfate) that are corrosive to the vitrifier or the downstream air pollution control equipment, especially under the elevated temperature conditions existent in these components. Removal of these corrosives presents a significant challenge because most tank wastes contain high (up to 10-molar) concentrations of sodium hydroxide (NaOH) along with significant levels of nitrate, nitrite, and other anions, which render standard ion-exchange, membrane filtration, and other separation technologies relatively ineffective. In Task 20, the Energy and Environmental Research Center (EERC) will work with Duratek to develop and optimize a vitrification pretreatment process for consistent, quantitative removal of chloride and sulfate prior to vitrifier injection

  19. EMERGING TECHNOLOGY SUMMARY: VITRIFICATION OF SOILS CONTAMINATED BY HAZARDOUS AND/OR RADIOACTIVE WASTES

    Science.gov (United States)

    A performance summary of an advanced multifuel-capable combustion and melting system (CMS) for the vitrification of hazardous wastes is presented. Vortex Corporation has evaluated its patented CMS for use in the remediation of soils contaminated with heavy metals and radionuclid...

  20. The feasibility of sampling the glass pour in a high level waste vitrification plant

    International Nuclear Information System (INIS)

    Cole, G.V.; Shilton, P.; Morris, J.B.

    1986-06-01

    Vitrified high level waste can be sampled for quality assurance purposes in three general ways: (I) from the glass pour, (II) from the canister, and (III) from the melter. A discussion of the potential advantages and disadvantages of each route is presented. The second philosophy seems to show the best promise; it is recommended that the Contained Pot method and the Token method are best suited for further development. An international survey of policy at vitrification plants shows that with one possible exception no glass sampling is intended and that quality is normally to be assured by control of the vitrification process. (author)

  1. Powder technological vitrification of simulated high-level waste

    International Nuclear Information System (INIS)

    Gahlert, S.

    1988-03-01

    High-level waste simulate from the reprocessing of light water reactor and fast breeder fuel was vitrified by powder technology. After denitration with formaldehyde, the simulated HLW is mixed with glass frit and simultaneously dried in an oil-heated mixer. After 'in-can calcination' for at least 24 hours at 850 or 950 K (depending on the type of waste and glass), the mixture is hot-pressed in-can for several hours at 920 or 1020 K respectively, at pressures between 0.4 and 1.0 MPa. The technology has been demonstrated inactively up to diameters of 30 cm. Leach resistance is significantly enhanced when compared to common borosilicate glasses by the utilization of glasses with higher silicon and aluminium content and lower sodium content. (orig.) [de

  2. Fifth in situ vitrification engineering-scale test of simulated INEL buried waste sites

    International Nuclear Information System (INIS)

    Bergsman, T.M.; Shade, J.W.; Farnsworth, R.K.

    1992-06-01

    In September 1990, an engineering-scale in situ vitrification (ISV) test was conducted on sealed canisters containing a combined mixture of buried waste materials expected to be present at the Idaho National Engineering Laboratory (INEL) Subsurface Disposal Area (SDA). The test was part of a Pacific Northwest Laboratory (PNL) program to assist INEL in treatability studies of the potential application of ISV to mixed transuranic wastes at the INEL SDA. The purpose of this test was to determine the effect of a close-packed layer of sealed containers on ISV processing performance. Specific objectives included determining (1) the effect of releases from sealed containers on hood plenum pressure and temperature, (2) the release pressure ad temperatures of the sealed canisters, (3) the relationships between canister depressurization and melt encapsulation, (4) the resulting glass and soil quality, (5) the potential effects of thermal transport due to a canister layer, (6) the effects on particle entrainment of differing angles of approach for the ISV melt front, and (7) the effects of these canisters on the volatilization of voltatile and semivolatile contaminants into the hood plenum

  3. In situ vitrification of buried waste: Containment issues and suppression systems

    International Nuclear Information System (INIS)

    Luey, J.; Powell, T.D.

    1992-03-01

    Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) are developing a remedial action technology for buried waste through the adaptation of the in situ vitrification (ISV) process. The ISV process is a thermal treatment process originally developed for the US Department of Energy (DOE) to stabilize soils contaminated with transuranic waste. ISV tests with buried waste forms have demonstrated that the processing of buried waste is more dynamic than the processing of soils. This paper will focus on the issue of containment of the gases released during the processing of buried waste and on engineered suppression systems to alleviate transient events associated with dynamic off-gassing from the ISV melt

  4. Modifying the rheological properties of melter feed for the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Blair, H.T.; McMakin, A.H.

    1986-03-01

    Selected high-level nuclear wastes from the Hanford Site may be vitrified in the future Hanford Waste Vitrification Plant (HWVP) by Rockwell Hanford Company, the contractor responsible for reprocessing and waste management at the Hanford Site. The Pacific Northwest Laboratory (PNL), is responsible for providing technical support for the HWVP. In this capacity, PNL performed rheological evaluations of simulated HWVP feed in order to determine which processing factors could be modified to best optimize the vitrification process. To accomplish this goal, a simulated HWVP feed was first created and characterized. Researchers then evaluated how the chemical and physical form of the glass-forming additives affected the rheological properties and melting behavior of melter feed prepared with the simulated HWVP feed. The effects of adding formic acid to the waste were also evaluated. Finally, the maximum melter feed concentration with acceptable rheological properties was determined

  5. Demonstrating compliance with WAPS 1.3 in the Hanford waste vitrification plant process

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, M.F.; Piepel, G.F.; Simpson, D.B.

    1996-03-01

    The high-level waste (HLW) vitrification plant at the Hanford Site was being designed to immobilize transuranic and high-level radioactive waste in borosilicate glass. This document describes the statistical procedure to be used in verifying compliance with requirements imposed by Section 1.3 of the Waste Acceptance Product Specifications (WAPS, USDOE 1993). WAPS 1.3 is a specification for ``product consistency,`` as measured by the Product Consistency Test (PCT, Jantzen 1992b), for each of three elements: lithium, sodium, and boron. Properties of a process batch and the resulting glass are largely determined by the composition of the feed material. Empirical models are being developed to estimate some property values, including PCT results, from data on feed composition. These models will be used in conjunction with measurements of feed composition to control the HLW vitrification process and product.

  6. Small-scale integrated demonstration of high-level radioactive waste processing and vitrification using actual SRP waste

    International Nuclear Information System (INIS)

    Woolsey, G.B.; Baumgarten, P.K.; Eibling, R.E.; Ferguson, R.B.

    1981-01-01

    A small-scale pilot plant for chemical processing and vitrification of actual high-level waste has been constructed at the Savannah River Laboratory (SRL). This fully integrated facility has been constructed in six shielded cells and has eight major unit operations. Equipment performance and processing characteristics of the unit operations are reported

  7. Development of a glass matrix for vitrification of sulphate bearing high level radioactive liquid waste

    International Nuclear Information System (INIS)

    Kaushik, C.P.; Mishra, R.K.; Thorat, Vidya; Ramchandran, M.; Amar Kumar; Ozarde, P.D.; Raj, Kanwar; Das, D.

    2004-07-01

    High level radioactive liquid waste (HLW) is generated during reprocessing of spent nuclear fuel. In the earlier reprocessing flow sheet ferrous sulphamate has been used for valancy adjustment of Pu from IV to III for effective separation. This has resulted in generation of HLW containing significance amount of sulphate. Internationally borosilicate glass matrix has been adopted for vitrification of HLW. The first Indian vitrification facility at Waste Immobilislition Plant (WIP), Tarapur a five component borosilicate matrix (SiO 2 :B 2 O 3 :Na 2 O : MnO : TiO 2 ) has been used for vitrification of waste. However at Trombay HLW contain significant amount of sulphate which is not compatible with standard borosilicate formulation. Extensive R and D efforts were made to develop a glass formulation which can accommodate sulphate and other constituents of HLW e.g., U, Al, Ca, etc. This report deals with development work of a glass formulations for immobilization of sulphate bearing waste. Different glass formulations were studied to evaluate the compatibility with respect to sulphate and other constituents as mentioned above. This includes sodium, lead and barium borosilicate glass matrices. Problems encountered in different glass matrices for containment of sulphate have also been addressed. A glass formulation based on barium borosilicate was found to be effective and compatible for sulphate bearing high level waste. (author)

  8. Design and operation of high level waste vitrification and storage facilities

    International Nuclear Information System (INIS)

    1992-01-01

    The conversion of high level wastes (HLW) into solids has been studied for the past 30 years, primarily in those countries engaged in the reprocessing of nuclear fuels. Production and demonstration calcination and solidification plants have been operated by using waste solutions from fuels irradiated at various burnup rates, depending on the reactor type. Construction of more advanced solidification processes is now in progress in several countries to permit the handling of high burnup power reactor fuel wastes. The object of this report is to provide detailed information and references for those vitrification systems in advanced stages of implementation. Some less detailed information will be provided for previously developed immobilization systems. The report will examine the HLLW arising from the various locations, the features of each process as well as the stage of development, scale-up potential and flexibility of the processes. Since the publication of IAEA Technical Reports Series No. 176, Techniques for the Solidification of High-Level Wastes great progress on this subject has been made. The AVM in France has been operated successfully for 11 years and France has completed construction at La Hague of two vitrification plants that are based on the AVM rotary calciner/metallic melter process. A similar plant is under construction at Sellafield. The ceramic melter process has been chosen by several countries. Germany has successfully operated the PAMELA vitrification plant. Since 1986, Belgoprocess has continued to operate this facility. The former USSR operated the EP-500 plant from 1986 to 1988. In addition, two ceramic melter vitrification plants are nearing completion in the USA at Savannah River and West Valley and plans are being made to use this technology at Hanford as well as in Japan, Germany and India. This major progress attests to the maturity of these technologies for vitrifying HLLW to make a borosilicate glass for disposal of the waste. 67

  9. Radioactive waste combustion-vitrification under arc plasma: thermal and dynamic modelling

    International Nuclear Information System (INIS)

    Barthelemy, B.

    2003-06-01

    This thesis concerns the thermal and dynamic modelling for a combustion/vitrification process of surrogate radioactive waste under transferred arc plasma. The writer presents the confinement processes for radioactive waste using arc plasma and the different software used to model theses processes. This is followed by a description of our experimental equipment including a plasma arc reactor and an inductive system allowing the homogenization of glass temperature. A combustion/vitrification test is described. Thermal and material balances were discussed. The temperature fields of plasma arc and the glass frit conductivity are measured. Finally, the writer describes and clarifies the equations solved for the simulations of the electrically plasma arc and the glass melting including the thin layer of glass frit coating the crucible cold walls. The modelling results are presented in the form of spatial distribution of temperature, velocity and voluminal power... (author)

  10. Radioactive waste combustion / vitrification under arc plasma: thermal and dynamic modelling

    International Nuclear Information System (INIS)

    Barthelemy, B.

    2003-01-01

    This thesis concerns the thermal and dynamic modelling for a combustion/vitrification process of surrogate radioactive waste under transferred arc plasma. The writer presents the confinement processes for radioactive waste using arc plasma and the different software used to model theses processes. This is followed by a description of our experimental equipment including a plasma arc reactor and an inductive system allowing the homogenization of glass temperature. A combustion/vitrification test is described. Thermal and material balances were discussed. The temperature fields of plasma arc and the glass frit conductivity are measured. Finally, the writer describes and clarifies the equations solved for the simulations of the electrically plasma arc and the glass melting including the thin layer of glass frit coating the crucible cold walls. The modelling results are presented in the form of spatial distribution of temperature, velocity and volume power... (author)

  11. High temperature materials for radioactive waste incineration and vitrification. Revision 1

    International Nuclear Information System (INIS)

    Bickford, D.F.; Ondrejcin, R.S.; Salley, L.

    1986-01-01

    Incineration or vitrification of radioactive waste subjects equipment to alkaline or acidic fluxing, oxidation, sulfidation, carburization, and thermal shock. It is necessary to select appropriate materials of construction and control operating conditions to avoid rapid equipment failure. Nickel- and cobalt-based alloys with high chromium or aluminum content and aluminum oxide/chromium oxide refractories with high chromium oxide content have provided the best service in pilot-scale melter tests. Inconel 690 and Monofrax K-3 are being used for waste vitrification. Haynes 188 and high alumina refractory are undergoing pilot scale tests for incineration equipment. Laboratory tests indicate that alloys and refractories containing still higher concentrations of chromium or chromium oxide, such as Inconel 671 and Monofrax E, may provide superior resistance to attack in glass melter environments

  12. Tolerancing requirements for remote handling at the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Van Katwijk, C.; Keenan, R.M.; Bullis, R.E.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) is being designed by Fluor Daniel, Inc. with Waste Chem Corporation as Fluor Daniel, Inc.'s major subcontractor specializing in vitrification and remote system technologies. United Engineers and Constructors (UE ampersand C)/Catalytic (UCAT) will construct the plant. Westinghouse Hanford Company is the Project Integration manager and Business manager, and as the plant operator it provides technical direction to the Architect/ Engineer team (A/E) and constructor on behalf of the US Department of Energy - Richland Field Office. The A/E has developed, in cooperation with UE ampersand C, Westinghouse Hanford Company, and the US Department of Energy, a new and innovative approach to installations of the many remote nozzles and electrical connectors that must be installed to demanding tolerances. This paper summarizes the key elements of the HWVP approach

  13. The hot bench scale plant Ester for the vitrification of high level wastes

    International Nuclear Information System (INIS)

    Nannicini, R.; Strazzer, A.; Cantale, C; Donato, A.; Grossi, G.

    1985-01-01

    In this paper the hot bench-scale plant ESTER for the vitrification of the high-level radioactive wastes is described, and the main results of the first radioactive campaign are reported. The ESTER plant, which is placed in the ADECO-ESSOR hot cells of the C.C.R.-EURATOM-ISPRA, has been built and is operated by the ENEA, Departement of Fuel Cycle. It began operating with real radioactive wastes about 1 year ago, solidifying a total of 12 Ci of fission products into 2,02 Kg of borosilicate glass, corresponding to 757 ml of glass. During the vitrification many samples of liquid and gaseous streams have been taken and analyzed. A radioactivity balance in the plant has been calculated, as well as a mass balance of nitrates and of the 137 Cs and 106 Ru volatized in the process

  14. Design features of a full-scale high-level waste vitrification system

    International Nuclear Information System (INIS)

    Siemens, D.H.; Bonner, W.F.

    1976-08-01

    A system has been designed and is currently under construction for vitrification of commercial high-level waste. The process consists of a spray calciner coupled to an in-can melter. Due to the high radiation levels expected, this equipment is designed for totally remote operation and maintenance. The in-cell arrangement of this equipment has been developed cooperatively with a nuclear fuel reprocessor. The system will be demonstrated both full scale with nonradioactive simulated waste and pilot scale with actual high-level waste

  15. Critique of Hanford Waste Vitrification Plant off-gas sampling requirements

    International Nuclear Information System (INIS)

    Goles, R.W.

    1996-03-01

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

  16. Thermal treatment and vitrification of boiler ash from a municipal solid waste incinerator.

    Science.gov (United States)

    Yang, Y; Xiao, Y; Voncken, J H L; Wilson, N

    2008-06-15

    Boiler ash generated from municipal solid waste (MSW) incinerators is usually classified as hazardous materials and requires special disposal. In the present study, the boiler ash was characterized for the chemical compositions, morphology and microstructure. The thermal chemical behavior during ash heating was investigated with thermal balance. Vitrification of the ash was conducted at a temperature of 1400 degrees C in order to generate a stable silicate slag, and the formed slag was examined with chemical and mineralogical analyses. The effect of vitrification on the leaching characteristics of various elements in the ash was evaluated with acid leaching. The study shows that the boiler ash as a heterogeneous fine powder contains mainly silicate, carbonate, sulfates, chlorides, and residues of organic materials and heavy metal compounds. At elevated temperatures, the boiler ash goes through the initial moisture removal, volatilization, decomposition, sintering, melting, and slag formation. At 1400 degrees C a thin layer of salt melt and a homogeneous glassy slag was formed. The experimental results indicate that leaching values of the vitrified slag are significantly reduced compared to the original boiler ash, and the vitrification could be an interesting alternative for a safer disposal of the boiler ash. Ash compacting, e.g., pelletizing can reduce volatilization and weight loss by about 50%, and would be a good option for the feed preparation before vitrification.

  17. Start-up of commercial high level waste vitrification facilities at La Hague

    International Nuclear Information System (INIS)

    Sombret, C.; Jouan, A.; Fournier, W.; Alexandre, D.; Leroy, L.

    1991-01-01

    The paper describes industial experience gained in France for vitrification of fission products generated by spent fuel reprocessing. The continuous vitrification process developed by CEA, SGN and COGEMA is outlined and Marcoule Vitrification Facility (AVM), with output results since start-up of hot operation in June 1978, briefly presented. Vitrification of high-level liquid waste has now entered an industrial phase at La Hague with R7 and T7 facilities. R7 and T7 have each been designed to process FP solutions generated by reprocessing LWR fuel with an initial enrichment of 3.5% and a discharge burn-up of 33,000 MWd/t. R7 active operations began on June, 1989. This facility is now vitrifying the backlog of fission products resulting from the existing UP2 reprocessing plant, which is being currently extended. Scheduled to start early in 1992, T7 will vitrify the fission products, dissolution fines and sodium-rich solutions issuing from UP3 plant

  18. The impact of safety analyses on the design of the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Koppenaal, T.J.; Yee, A.K.; Reisdorf, J.B.; Hall, B.W.

    1993-04-01

    Accident analyses are being performed to evaluate and document the safety of the Hanford Waste Vitrification Plant (HWVP). The safety of the HWVP is assessed by evaluating worst-case accident scenarios and determining the dose to offsite and onsite receptors. Air dispersion modeling is done with the GENII computer code. Three accidents are summarized in this paper, and their effects on the safety and the design of the HWVP are demonstrated

  19. Steady state simulation of Joule heated ceramic melter for vitrification of high level liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Sugilal, G; Wattal, P K; Theyyunni, T K [Process Engineering and Systems Development Division, Bhabha Atomic Research Centre, Mumbai (India); Iyer, K N [Department of Mechanical Engineering, Indian Inst. of Tech., Mumbai (India)

    1994-06-01

    The Joule heated ceramic melter is emerging as an attractive alternative to metallic melters for high level waste vitrification. The inherent limitations with metallic melters viz., low capacity and short melter life, are overcome in a ceramic melter which can be adopted for continuous mode of operation. The ceramic melter has the added advantage of better operational flexibility. This paper describes the three dimensional model used for simulating the complex design conditions of the ceramic melter. (author).

  20. Removal of Aerosol Particles Generated from Vitrification Process for High-Level Liquid Wastes

    OpenAIRE

    加藤 功

    1990-01-01

    The vitrification technology has been developed for the high-level liquid waste (HLLW) from reprocessing nuclear spent fuel in PNC. The removal performance of the aerosol particles generated from the melting process was studied in a nonradioactive full-scale mock-up test facility (MTF). The off-gas treatment system consists of submerged bed scrubber (SBS), venturi scrubber, NOx absorber, high efficiency mist eliminater (HEME). Deoomtamination factors (DFs) were derived from the mass ratio of ...

  1. Hanford Waste Vitrification Plant technical background document for toxics best available control technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    None

    1992-10-01

    This document provides information on toxic air pollutant emissions to support the Notice of Construction for the proposed Hanford Waste Vitrification Plant (HWVP) to be built at the the Department of Energy Hanford Site near Richland, Washington. Because approval must be received prior to initiating construction of the facility, state and federal Clean Air Act Notices of construction are being prepared along with necessary support documentation.

  2. Steady state simulation of Joule heated ceramic melter for vitrification of high level liquid waste

    International Nuclear Information System (INIS)

    Sugilal, G.; Wattal, P.K.; Theyyunni, T.K.; Iyer, K.N.

    1994-01-01

    The Joule heated ceramic melter is emerging as an attractive alternative to metallic melters for high level waste vitrification. The inherent limitations with metallic melters viz., low capacity and short melter life, are overcome in a ceramic melter which can be adopted for continuous mode of operation. The ceramic melter has the added advantage of better operational flexibility. This paper describes the three dimensional model used for simulating the complex design conditions of the ceramic melter. (author)

  3. A state of the art review of vitrification of high level waste in Europe

    International Nuclear Information System (INIS)

    Heimerl, W.

    This paper gives a review of the state-of-the-art of the development and demonstration of vitrification processes for high level radioactive waste solutions on an industrial scale in four European countries (France, United Kingdom, Federal Republic of Germany and Belgium). Historical development, experiences and operations present status and future plans are presented. Three of the processes which seem to be of major importance are described (AVM-France, HARVEST-United Kingdom, PAMELA Germany/Belgium). (author)

  4. Hanford Waste Vitrification Plant technical background document for toxics best available control technology demonstration

    International Nuclear Information System (INIS)

    1992-10-01

    This document provides information on toxic air pollutant emissions to support the Notice of Construction for the proposed Hanford Waste Vitrification Plant (HWVP) to be built at the the Department of Energy Hanford Site near Richland, Washington. Because approval must be received prior to initiating construction of the facility, state and federal Clean Air Act Notices of construction are being prepared along with necessary support documentation

  5. Los Alamos National Laboratory simulated sludge vitrification demonstration

    International Nuclear Information System (INIS)

    Cicero, C.A.; Bickford, D.F.; Bennert, D.M.; Overcamp, T.J.

    1994-01-01

    Technologies are being developed to convert hazardous and mixed wastes to a form suitable for permanent disposal. Vitrification, which has been declared the Best Demonstrated Available Technology (BDAT) for high-level radioactive waste disposal by the EPA, is capable of producing a highly durable wasteform that minimizes disposal volumes through organic destruction, moisture evaporation, and porosity reduction. However, this technology must be demonstrated over a range of waste characteristics, including compositions, chemistries, moistures, and physical characteristics to ensure that it is suitable for hazardous and mixed waste treatment. This project plans to demonstrate vitrification of simulated wastes that are considered representatives of wastes found throughout the DOE complex. For the most part, the primary constituent of the wastes is flocculation aids, such as Fe(OH) 3 , and natural filter aids, such as diatomaceous earth and perlite. The filter aids consist mostly of silica, which serves as an excellent glass former; hence, the reason why vitrification is such a viable option. LANL is currently operating a liquid waste processing plant which produces an inorganic sludge similar to other waste water treatment streams. Since this waste has characteristics that make it suitable for vitrification and the likelihood of success is high, it shall be tested at CU. The objective of this task is to characterize the process behavior and glass product formed upon vitrification of simulated LANL sludge. The off-gases generated from the production runs will also be characterized to help further develop vitrification processes for mixed and low level wastes

  6. Feasibility study on vitrification of low- and intermediate-level radioactive waste from pressurized water reactors

    International Nuclear Information System (INIS)

    Park, J.K.; Song, M.J.

    1998-01-01

    In order to obtain annual generation volume and composition data for low- and intermediate-level radioactive waste (LILW), characteristics and generation trends for each waste which was produced at nuclear power plants (NPPs) in Korea were investigated. Of the three different types of melters, the platinum crucible was found to be most suitable for the performance of vitrification experiments and hence, was used to help better understand the optimal waste contents in borosilicate glass waste forms with respect to waste types. After the performance of vitrification experiments, compressive strength tests showed that the final waste glass product, containing up to 40 vol% of ashy pyrolyzed/oxidized at 400--800 C, showed good mechanical stability and homogeneity in the glass matrix. Economical assessment was performed with some considerations given for equipment having already been adopted for LILW treatment in Korea for four treatment strategies with melters selected from a technical assessment. For each strategy, the capital and the operation cost were estimated, and the disposal volume was calculated with reasonably estimated volume reduction factors with regard to waste type and treatment concept

  7. Summary Of Cold Crucible Vitrification Tests Results With Savannah River Site High Level Waste Surrogates

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

    2014-01-13

    The cold crucible inductive melting (CCIM) technology successfully applied for vitrification of low- and intermediate-level waste (LILW) at SIA Radon, Russia, was tested to be implemented for vitrification of high-level waste (HLW) stored at Savannah River Site, USA. Mixtures of Sludge Batch 2 (SB2) and 4 (SB4) waste surrogates and borosilicate frits as slurries were vitrified in bench- (236 mm inner diameter) and full-scale (418 mm inner diameter) cold crucibles. Various process conditions were tested and major process variables were determined. Melts were poured into 10L canisters and cooled to room temperature in air or in heat-insulated boxes by a regime similar to Canister Centerline Cooling (CCC) used at DWPF. The products with waste loading from ~40 to ~65 wt.% were investigated in details. The products contained 40 to 55 wt.% waste oxides were predominantly amorphous; at higher waste loadings (WL) spinel structure phases and nepheline were present. Normalized release values for Li, B, Na, and Si determined by PCT procedure remain lower than those from EA glass at waste loadings of up to 60 wt.%.

  8. An example of toxic effluents treatment: nuclear waste vitrification

    International Nuclear Information System (INIS)

    Moncouyoux, J.P.; Teulon, F.

    1993-01-01

    The management of radioactive waste has to aim to avoid to bring nuisances to future generations whom industrial people, public authorities would be responsible. The French way of doing in radioactive waste storage is called multi-barriers, to forbid or limit radionuclides migration to the biosphere. This containment system is composed of three barriers: the waste packaging, the storage and the site or geological barrier. This way of doing can also be used for toxic industrial waste. 4 refs

  9. Efficient handling of high-level radioactive cell waste in a vitrification facility analytical laboratory

    International Nuclear Information System (INIS)

    Roberts, D.W.; Collins, K.J.

    1998-01-01

    The Savannah River Site''s (SRS) Defense Waste Processing Facility (DWPF) near Aiken, South Carolina, is the world''s largest and the United State''s first high level waste vitrification facility. For the past 1.5 years, DWPF has been vitrifying high level radioactive liquid waste left over from the Cold War. The vitrification process involves the stabilization of high level radioactive liquid waste into borosilicate glass. The glass is contained in stainless steel canisters. DWPF has filled more than 200 canisters 3.05 meters (10 feet) long and 0.61 meters (2 foot) diameter. Since operations began at DWPF in March of 1996, high level radioactive solid waste continues to be generated due to operating the facility''s analytical laboratory. The waste is referred to as cell waste and is routinely removed from the analytical laboratories. Through facility design, engineering controls, and administrative controls, DWPF has established efficient methods of handling the high level waste generated in its laboratory facility. These methods have resulted in the prevention of undue radiation exposure, wasted man-hours, expenses due to waste disposal, and the spread of contamination. This level of efficiency was not reached overnight, but it involved the collaboration of Radiological Control Operations and Laboratory personnel working together to devise methods that best benefited the facility. This paper discusses the methods that have been incorporated at DWPF for the handling of cell waste. The objective of this paper is to provide insight to good radiological and safety practices that were incorporated to handle high level radioactive waste in a laboratory setting

  10. Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system

    International Nuclear Information System (INIS)

    Powell, J.; Reich, M.; Barletta, R.

    1996-01-01

    A new approach, termed SMILE (Small Module Inductively Loaded Energy), for the vitrification of high level nuclear wastes (HLW) is described. Present vitrification systems liquefy the HLW solids and associated frit material in large high temperature melters. The molten mix is then poured into small (∼1 m 3 ) disposal canisters, where it solidifies and cools. SMILE eliminates the separate, large high temperature melter. Instead, the BLW solids and frit melt inside the final disposal containers, using inductive heating. The contents then solidify and cool in place. The SMILE modules and the inductive heating process are designed so that the outer stainless can of the module remains at near ambient temperature during the process cycle. Module dimensions are similar to those of present disposal containers. The can is thermally insulated from the high temperature inner container by a thin layer of refractory alumina firebricks. The inner container is a graphite crucible lined with a dense alumina refractory that holds the HLW and fiit materials. After the SMILE module is loaded with a slurry of HLW and frit solids, an external multi-turn coil is energized with 30-cycle AC current. The enclosing external coil is the primary of a power transformer, with the graphite crucible acting as a single turn ''secondary.'' The induced current in the ''secondary'' heats the graphite, which in turn heats the HLW and frit materials. The first stage of the heating process is carried out at an intermediate temperature to drive off remnant liquid water and water of hydration, which takes about 1 day. The small fill/vent tube to the module is then sealed off and the interior temperature raised to the vitrification range, i.e., ∼1200C. Liquefaction is complete after approximately 1 day. The inductive heating then ceases and the module slowly loses heat to the environment, allowing the molten material to solidify and cool down to ambient temperature

  11. In situ vitrification: A new process for waste remediation

    International Nuclear Information System (INIS)

    Fitzpatrick, V.F.; Timmerman, C.L.; Buelt, J.L.

    1987-07-01

    In situ vitrification is a thermal treatment process that converts contaminated soil into a chemically inert, stable glass and crystalline product. A square array of four electrodes are inserted into the ground to the desired treatment depth. Because the soil is not electrically conductive once the moisture has been driven off, a conductive mixture of flaked graphite and glass frit is placed among the electrodes to act as the starter path. An electrical potential is applied to the electrodes, which establishes an electrical current in the starter path. The resultant power heats the starter path and surrounding soil up to 3600 0 F, well above the normal fusion temperature of soil of between 2000 and 2500 0 F. The graphite starter path is eventually consumed by oxidation, and the current is transferred to the molten soil, which is now electrically conductive. As the vitrified zone grows, it incorporates nonvolatile elements and destroys organic components by pyrolysis. The pyrolyzed byproducts migrate to the surface of the vitrified zone, where they combust in the presence of oxygen. A hood placed over the processing area provides confinement for the combustion gases, and the gases are drawn into the off-gas treatment system. 8 refs., 7 figs., 2 tabs

  12. High-temperature vitrification of low-level radioactive and hazardous wastes

    International Nuclear Information System (INIS)

    Schumacher, R.F.; Kielpinski, A.L.; Bickford, D.F.; Cicero, C.A.; Applewhite-Ramsey, A.; Spatz, T.L.; Marra, J.C.

    1995-01-01

    The US Department of Energy (DOE) weapons complex has numerous radioactive waste streams which cannot be easily treated with joule-heated vitrification systems. However, it appears these streams could be treated With certain robust, high-temperature, melter technologies. These technologies are based on the use of plasma torch, graphite arc, and induction heating sources. The Savannah River Technology Center (SRTC), with financial support from the Department of Energy, Office of Technology Development (OTD) and in conjunction with the sites within the DOE weapons complex, has been investigating high-temperature vitrification technologies for several years. This program has been a cooperative effort between a number of nearby Universities, specific sites within the DOE complex, commercial equipment suppliers and the All-Russian Research Institute of Chemical Technology. These robust vitrification systems appear to have advantages for the waste streams containing inorganic materials in combination with significant quantities of metals, organics, salts, or high temperature materials. Several high-temperature technologies were selected and will be evaluated and employed to develop supporting technology. A general overview of the SRTC ''High-Temperature Program'' will be provided

  13. High-temperature vitrification of low-level radioactive and hazardous wastes

    International Nuclear Information System (INIS)

    Schumacher, R.F.; Kielpinski, A.L.; Bickford, D.F.; Cicero, C.A.; Applewhite-Ramsey, A.; Spatz, T.L.; Marra, J.C.

    1995-01-01

    The US Department of Energy (DOE) weapons complex has numerous radioactive waste streams which cannot be easily treated with joule-heated vitrification systems. However, it appears that these streams could be treated with certain robust, high-temperature, melter technologies. These technologies are based on the use of plasma torch, graphite arc, and induction heating sources. The Savannah River Technology Center (SRTC), with financial support from the Department of Energy, Office of Technology Development (OTD) and in conjunction with the sites within the DOE weapons complex, has been investigating high-temperature vitrification technologies for several years. This program has been a cooperative effort between a number of nearby Universities, specific sites within the DOE complex, commercial equipment suppliers and the All-Russian Research Institute of Chemical Technology. These robust vitrification systems appear to have advantages for the waste streams containing inorganic materials in combination with significant quantities of metals, organics, salts, or high temperature materials. Several high-temperature technologies were selected and will be evaluated and employed to develop supporting technology. A general overview of the SRTC ''High-Temperature Program'' will be provided

  14. Strategy for product composition control in the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Bryan, M.F.; Piepel, G.F.

    1996-03-01

    The Hanford Waste Vitrification Plant (HWVP) will immobilize transuranic and high-level radioactive waste in borosilicate glass. The major objective of the Process/Product Model Development (PPMD) cost account of the Pacific Northwest Laboratory HWVP Technology Development (PHTD) Project is the development of a system for guiding control of feed slurry composition (which affects glass properties) and for checking and documenting product quality. This document lays out the broad structure of HWVP's product composition control system, discusses five major algorithms and technical issues relevant to this system, and sketches the path of development and testing

  15. Modeling of NOx Destruction Options for INEEL Sodium-Bearing Waste Vitrification

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Richard Arthur

    2001-09-01

    Off-gas NOx concentrations in the range of 1-5 mol% are expected as a result of the proposed vitrification of sodium-bearing waste at the Idaho National Engineering and Environmental Laboratory. An existing kinetic model for staged combustion (originally developed for NOx abatement from the calcination process) was updated for application to vitrification offgas. In addition, two new kinetic models were developed to assess the feasibility of using selective non-catalytic reduction (SNCR) or high-temperature alone for NOx abatement. Each of the models was developed using the Chemkin code. Results indicate that SNCR is a viable option, reducing NOx levels to below 1000 ppmv. In addition, SNCR may be capable of simultaneously reducing CO emissions to below 100 ppmv. Results for using high-temperature alone were not as promising, indicating that a minimum NOx concentration of 3950 ppmv is achievable at 3344°F.

  16. Overview of mixed waste issues

    International Nuclear Information System (INIS)

    Piciulo, P.L.; Bowerman, B.S.; Kempf, C.R.; MacKenzie, D.R.; Siskind, B.

    1986-01-01

    Based on BNL's study it was concluded that there are LLWs which contain chemically hazardous components. Scintillation liquids may be considered an EPA listed hazardous waste and are, therefore, potential mixed wastes. Since November, 1985 no operating LLW disposal site will accept these wastes for disposal. Unless such wastes contain de minimis quantities of radionuclides, they cannot be disposed of at an EPA an EPA permitted site. Currently generators of LSC wastes can ship de minimis wastes to be burned at commercial facilities. Oil wastes will also eventually be an EPA listed waste and thus will have to be considered a potential radioactive mixed wasted unless NRC establishes de minimis levels of radionuclides below which oils can be managed as hazardous wastes. Regarding wastes containing lead metal there is some question as to the extent of the hazard posed by lead disposed in a LLW burial trench. Chromium-containing wastes would have to be tested to determine whether they are potential mixed wastes. There may be other wastes that are mixed wastes; the responsibility for determining this rests with the waste generator. It is believed that there are management options for handling potential mixed wastes but there is no regulatory guidance. BNL has identified and evaluated a variety of treatment options for the management of potential radioactive mixed wastes. The findings of that study showed that application of a management option with the purpose of addressing EPA concern can, at the same time, address stabilization and volume reduction concerns of NRC

  17. Subsurface Planar Vitrification Treatment of Problematic TRU Wastes: Status of a Technology Demonstration Program

    International Nuclear Information System (INIS)

    Morse, M.K.; Nowack, B.R.; Thompson, L.E.

    2006-01-01

    This paper provides a status of the In Situ Transuranic Waste Delineation and Removal Project in which the GeoMelt R Subsurface Planar Vitrification TM (SPV TM ) process is being evaluated for the in situ treatment of burial sites containing remote handled mixed transuranic (TRU) waste. The GeoMelt R SPV TM process was invented and patented by Geosafe Corporation. AMEC holds the exclusive worldwide license to use this technology. The current project is part of a three-phase demonstration program to evaluate the effectiveness of the GeoMelt R SPV TM process to treat waste contained in vertical pipe units (VPUs) and caissons that were used for the disposal of remote handled transuranic wastes located at Hanford's 618-10 and 618-11 burial grounds. This project is being performed for the US Department of Energy (DOE) for use at the Hanford site and other DOE installations. The Phase I evaluation determined that removal and treatment of the 618-10/11 VPUs are beyond what can be safely accomplished using conventional excavation methods. Accordingly, a careful stepwise non-intrusive delineation approach and treatment using the GeoMelt R SPV TM technology, followed by removal, characterization, and disposal of the resulting inert vitrified mass was identified as the preferred alternative. Phase II of the project, which started in July 2004, included a full-scale non-radioactive demonstration of AMEC's GeoMelt R SPV TM process on a mock VPU configured to match the actual VPUs. The non-radioactive demonstration (completed in May 2005) was performed to confirm the approach and design before proceeding to a radioactive ('hot') demonstration on an actual VPU. This demonstration took approximately 130 hours, processed the entire mock VPU, and resulted in a vitrified monolith weighing an estimated 90 tonnes. [1] Plans for a radioactive demonstration on an actual VPU are being developed for CY 2006. In addition to demonstrating GeoMelt R SPV TM , delineation techniques are being

  18. Non-combustible waste vitrification with plasma torch melter.

    Science.gov (United States)

    Park, J K; Moon, Y P; Park, B C; Song, M J; Ko, K S; Cho, J M

    2001-05-01

    Non-combustible radioactive wastes generated from Nuclear Power Plants (NPPs) are composed of concrete, glass, asbestos, metal, sand, soil, spent filters, etc. The melting tests for concrete, glass, sand, and spent filters were carried out using a 60 kW plasma torch system. The surrogate wastes were prepared for the tests. Non-radioactive Co and Cs were added to the surrogates in order to simulate the radioactive waste. Several kinds of surrogate prepared by their own mixture or by single waste were melted with the plasma torch system to produce glassy waste forms. The characteristics of glassy waste forms were examined for the volume reduction factor (VRF) and the leach rate. The VRFs were estimated through the density measurement of the surrogates and the glassy waste forms, and were turned out to be 1.2-2.4. The EPA (Environmental Protection Agency) Toxicity Characteristic Leaching Procedure (TCLP) was used to determine the leach resistance for As, Ba, Hg, Pb, Cd, Cr, Se, Co, and Cs. The leaching index was calculated using the total content of each element in both the waste forms and the leachant. The TCLP tests resulted in that the leach rates for all elements except Co and Cs were lower than those of the Universal Treatment Standard (UTS) limits. There were no UTS limits for Co and Cs, and their leach rate & index from the experiments were resulted in around 10 times higher than those of other elements.

  19. In situ vitrification of buried waste: Containment issues and suppression systems

    International Nuclear Information System (INIS)

    Luey, J.; Powell, T.D.

    1992-01-01

    Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) are developing a remedial action technology for buried waste through the adaptation of the in situ vitrification (ISV) process. The ISV process is a thermal treatment process originally developed for the U.S. Department of Energy (DOE) to stabilize soils contaminated with transuranic waste. ISV tests with buried waste forms have demonstrated that the processing of buried waste is more dynamic than the processing of soils. This paper will focus on the issue of containment of the gases released during the processing of buried waste and on engineered suppression systems to alleviate transient events associated with dynamic off-gassing from the ISV melt. (author)

  20. Vitrification of high-level radioactive and hazardous wastes

    International Nuclear Information System (INIS)

    Lutze, W.

    1993-12-01

    The main objective is to summarize work conducted on glasses as waste forms for high-level radioactive fission product solutions up to the late 1980's (section I and II). Section III addresses the question, whether waste forms designed for the immobilization of radioactive residues can be used for the same purpose for hazardous wastes. Of particular interest are those types of hazardous wastes, e.g., fly ashes from municipal combustion plants, easy to convert into glasses or ceramic materials. A large number of base glass compositions has been studied to vitrify waste from reprocessing but only borosilicate glasses with melting temperatures between 1100 C and 1200 C and very good hydrolytic stability is used today. (orig./HP) [de

  1. Environmental Management vitrification activities

    Energy Technology Data Exchange (ETDEWEB)

    Krumrine, P.H. [Waste Policy Institute, Gaithersburg, MD (United States)

    1996-05-01

    Both the Mixed Waste and Landfill Stabilization Focus Areas as part of the Office of Technology Development efforts within the Department of Energy`s (DOE) Environmental Management (EM) Division have been developing various vitrification technologies as a treatment approach for the large quantities of transuranic (TRU), TRU mixed and Mixed Low Level Wastes that are stored in either landfills or above ground storage facilities. The technologies being developed include joule heated, plasma torch, plasma arc, induction, microwave, combustion, molten metal, and in situ methods. There are related efforts going into development glass, ceramic, and slag waste form windows of opportunity for the diverse quantities of heterogeneous wastes needing treatment. These studies look at both processing parameters, and long term performance parameters as a function of composition to assure that developed technologies have the right chemistry for success.

  2. Environmental Management vitrification activities

    International Nuclear Information System (INIS)

    Krumrine, P.H.

    1996-01-01

    Both the Mixed Waste and Landfill Stabilization Focus Areas as part of the Office of Technology Development efforts within the Department of Energy's (DOE) Environmental Management (EM) Division have been developing various vitrification technologies as a treatment approach for the large quantities of transuranic (TRU), TRU mixed and Mixed Low Level Wastes that are stored in either landfills or above ground storage facilities. The technologies being developed include joule heated, plasma torch, plasma arc, induction, microwave, combustion, molten metal, and in situ methods. There are related efforts going into development glass, ceramic, and slag waste form windows of opportunity for the diverse quantities of heterogeneous wastes needing treatment. These studies look at both processing parameters, and long term performance parameters as a function of composition to assure that developed technologies have the right chemistry for success

  3. Alternatives to High-Level Waste Vitrification: The Need for Common Sense

    International Nuclear Information System (INIS)

    Bell, Jimmy T.

    2000-01-01

    The competition for government funding for remediation of defense wastes (and for other legitimate government functions) is intensifying as the United States moves toward a balanced national budget. Determining waste remediation priorities for the use of available tax dollars will likely depend on established international agreements and on the real risks posed to human health.Remediation of the U.S. Department of Energy (DOE) high-level radioactive tank wastes has been described as the most important priority in the DOE system. The proposed tank waste remediation at three DOE sites will include retrieval of the wastes from the aging storage tanks, immobilization of the wastes, and safe disposal of the processed waste. Vitrification, the current immobilization technology chosen by DOE, is very costly. The U.S. Congress and the American people may not be aware that the present cost of preparing just 1 m 3 of processed waste product at the Savannah River Site is ∼$2 million! In a smaller waste remediation project at the West Valley Site, similar waste treatment is costing >$2 million/m 3 of waste product. Privatization efforts at the Hanford Site are now estimated to cost >$4 million/m 3 of waste product. Even at the lowest current cost of $2 million/m 3 of HLW glass product, the total estimated costs for remediating the tank wastes at the three DOE sites of Savannah River, Hanford, and Idaho Falls is $75 billion.Whether our nation can afford treatment costs of this magnitude and whether Congress will be willing to appropriate these huge sums for waste vitrification when alternative technologies can provide safe disposal at considerably lower cost are questions that need to be addressed. The hazard levels posed by the DOE tank wastes do not warrant high priority in comparison to the hazards of other defense wastes. Unless DOE selects a lower-cost technology for tank waste remediation, such efforts are likely to continue in a holding pattern, with little actually

  4. Alternatives to high-level waste vitrification: The need for common sense

    International Nuclear Information System (INIS)

    Bell, J.T.

    2000-01-01

    The competition for government funding for remediation of defense wastes (and for other legitimate government functions) is intensifying as the United states moves toward a balanced national budget. Determining waste remediation priorities for the use of available tax dollars will likely depend on established international agreements and on the real risks posed to human health. Remediation of the US Department of Energy (DOE) high-level radioactive tank wastes has been described as the most important priority in the DOE system. The proposed tank waste remediation at three DOE sites will include retrieval of the wastes from the aging storage tanks, immobilization of the wastes, and safe disposal of the processed waste. Vitrification, the current immobilization technology chosen by DOE, is very costly. The US Congress and the American people may not be aware that the present cost of preparing just 1 m 3 of processed waste product at the Savannah River Site is approximately$2 million. In a smaller waste remediation project at the West Valley Site, similar waste treatment is costing $2 million/m 3 of waste product. Privatization efforts at the Hanford Site are now estimated to cost $4 million/m 3 of waste product. Even at the lowest current cost of $2 million/m 3 of HLW glass product, the total estimated costs for remediating the tank wastes at the three DOE sites of Savannah River, Hanford, and Idaho Falls is $75 billion. Whether the nation can afford treatment costs of this magnitude and whether Congress will be willing to appropriate these huge sums for waste vitrification when alternative technologies can provide safe disposal at considerably lower cost are questions that need to be addressed. The hazard levels posed by the DOE tank wastes do not warrant high priority in comparison to the hazards of other defense wastes. Unless DOE selects a lower-cost technology for tank waste remediation, such efforts are likely to continue in a holding pattern, with little

  5. A feasibility study on the vitrification of low-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Song, Myung Jae; Park, Jong Kil; Ahn, Hee Jin [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Cho, Jeong Mi; Choe, Young Son; Cho, Myeong Ryul [Hankuk Fiber Group (Korea, Republic of)

    1995-12-31

    A Study was carried out to investigate the feasibility of vitrification for low-and medium-level radioactive waste(LMLW). In order to understand maximum yearly generation volume and composition for each waste streams waste generation trends, which have been produced from nuclear power plants(PWR) in korea, were examined and then technical and economical assessment were performed based on the volume and composition. To select the most promising melters, technical characteristics were analyzed for several melters such as cold crucible melter heated by direct induction(CCM), cold crucible melter heated by vertical electrodes(CCVE), molten metal melter(MM), and plasma melter(PM) which were most likely to be applied to LMLW treatment. Economical assessment was carried out for several treatment strategies with selected melters and resulted in that it was desirable that non-combustibles and spent filter were vitrified with PM, and the others with CCM. For the demonstration of vitrification possibility for protective clothing, vinyl seats, spent resin, and evaporator bottoms, the surrogated wastes were paralyzed or dried at optimal conditions and than specimens contained various percentages of pyrolysis ash were prepared with lab. and pilot scale melters. Compressive strength for these specimens was measured to determine the maximum ash content in glass waste forms. (author). 27 refs., 88 figs.

  6. A feasibility study on the vitrification of low-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Song, Myung Jae; Park, Jong Kil; Ahn, Hee Jin [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Cho, Jeong Mi; Choe, Young Son; Cho, Myeong Ryul [Hankuk Fiber Group (Korea, Republic of)

    1996-12-31

    A Study was carried out to investigate the feasibility of vitrification for low-and medium-level radioactive waste(LMLW). In order to understand maximum yearly generation volume and composition for each waste streams waste generation trends, which have been produced from nuclear power plants(PWR) in korea, were examined and then technical and economical assessment were performed based on the volume and composition. To select the most promising melters, technical characteristics were analyzed for several melters such as cold crucible melter heated by direct induction(CCM), cold crucible melter heated by vertical electrodes(CCVE), molten metal melter(MM), and plasma melter(PM) which were most likely to be applied to LMLW treatment. Economical assessment was carried out for several treatment strategies with selected melters and resulted in that it was desirable that non-combustibles and spent filter were vitrified with PM, and the others with CCM. For the demonstration of vitrification possibility for protective clothing, vinyl seats, spent resin, and evaporator bottoms, the surrogated wastes were paralyzed or dried at optimal conditions and than specimens contained various percentages of pyrolysis ash were prepared with lab. and pilot scale melters. Compressive strength for these specimens was measured to determine the maximum ash content in glass waste forms. (author). 27 refs., 88 figs.

  7. Regulatory aspects of mixed waste

    International Nuclear Information System (INIS)

    Boyle, R.R.; Orlando, D.A.

    1990-01-01

    Mixed waste is waste that satisfies the definition of low-level radioactive waste in the Low-Level Radioactive Waste Policy Amendments Act of 1985 (LLRWPAA) and contains hazardous waste that is either: (1) listed as a hazardous waste in 40 CFR 261, Subpart D; or (2) causes the waste to exhibit any of the characteristics identified in 40 CFR 261, Subpart C. Low-level radioactive waste is defined in the LLRWPAA as radioactive material that is not high level waste, spent nuclear fuel, or byproduct material, as defined in Section 11e(2) of the Atomic Energy Act of 1954, and is classified as low-level waste by the U.S. Nuclear Regulatory Commission (NRC). This paper discusses dual regulatory (NRC and Environmental Protection Agency) responsibility, overview of joint NRC/EPA guidance, workshops, national mixed waste survey, and principal mixed waste uncertainties

  8. A plasma melting of noncombustible waste for vitrification

    International Nuclear Information System (INIS)

    Moon, Young Pyo; Cho, Chun Hyung; Song, Myung Jae; Han, Sang Ok

    1997-01-01

    Multi-stage experiments have been under way to develop a new technology for radioactive waste treatment to reduce volume. Korea Electric Power Research Institute(KEPRI) has been seeking various technologies in order to reduce the radioactive volume significantly and to produce very stable waste forms. Plasma melting technology which offers greater control of temperature, faster time of reaction, better control of processing, lower capital costs, greater throughput, and more efficient use of energy has caught KEPRI's attention to the noncombustible radioactive waste. For the study of plasma melting for noncombustible wastes, KEPRI leased a lab scale multi-purpose plasma furnace together with accessory facilities and performed preliminary tests. The lab scale melting experiments were carried out by using the simulated noncombustible wastes based on the field survey data from nuclear power plants. KEPRI's current study is focused to find an optimum composition ratio of various noncombustible wastes for easy melting, to investigate physical properties of molten slag, and to obtain operating parameters for continuous operations

  9. Baseline tests for arc melter vitrification of INEL buried wastes. Volume 1: Facility description and summary data report

    International Nuclear Information System (INIS)

    Oden, L.L.; O'Connor, W.K.; Turner, P.C.; Soelberg, N.R.; Anderson, G.L.

    1993-01-01

    This report presents field results and raw data from the Buried Waste Integrated Demonstration (BWID) Arc Melter Vitrification Project Phase 1 baseline test series conducted by the Idaho National Engineering Laboratory (INEL) in cooperation with the U.S. Bureau of Mines (USBM). The baseline test series was conducted using the electric arc melter facility at the USBM Albany Research Center in Albany, Oregon. Five different surrogate waste feed mixtures were tested that simulated thermally-oxidized, buried, TRU-contaminated, mixed wastes and soils present at the INEL. The USBM Arc Furnace Integrated Waste Processing Test Facility includes a continuous feed system, the arc melting furnace, an offgas control system, and utilities. The melter is a sealed, 3-phase alternating current (ac) furnace approximately 2 m high and 1.3 m wide. The furnace has a capacity of 1 metric ton of steel and can process as much as 1,500 lb/h of soil-type waste materials. The surrogate feed materials included five mixtures designed to simulate incinerated TRU-contaminated buried waste materials mixed with INEL soil. Process samples, melter system operations data and offgas composition data were obtained during the baseline tests to evaluate the melter performance and meet test objectives. Samples and data gathered during this program included (a) automatically and manually logged melter systems operations data, (b) process samples of slag, metal and fume solids, and (c) offgas composition, temperature, velocity, flowrate, moisture content, particulate loading and metals content. This report consists of 2 volumes: Volume I summarizes the baseline test operations. It includes an executive summary, system and facility description, review of the surrogate waste mixtures, and a description of the baseline test activities, measurements, and sample collection. Volume II contains the raw test data and sample analyses from samples collected during the baseline tests

  10. High-temperature vitrification of Hanford residual-liquid waste in a continuous melter

    International Nuclear Information System (INIS)

    Barnes, S.M.

    1980-04-01

    Over 270 kg of high-temperature borosilicate glass have been produced in a series of three short-term tests in the High-Temperature Ceramic Melter vitrification system at PNL. The glass produced was formulated to vitrify simulated Hanford residual-liquid waste. The tests were designed to (1) demonstrate the feasibility of utilizing high-temperature, continuous-vitrification technology for the immobilization of the residual-liquid waste, (2) test the airlift draining technique utilized by the high-temperature melter, (3) compare glass produced in this process to residual-liquid glass produced under laboratory conditions, (4) investigate cesium volatility from the melter during waste processing, and (5) determine the maximum residual-liquid glass production rate in the high-temperature melter. The three tests with the residual-liquid composition confirmed the viability of the continuous-melting vitrification technique for the immobilization of this waste. The airlift draining technique was demonstrated in these tests and the glass produced from the melter was shown to be less porous than the laboratory-produced glass. The final glass produced from the second test was compared to a glass of the same composition produced under laboratory conditions. The comparative tests found the glasses to be indistinguishable, as the small differences in the test results fell within the precision range of the characterization testing equipment. The cesium volatility was examined in the final test. This examination showed that 0.44 wt % of the cesium (assumed to be cesium oxide) was volatilized, which translates to a volatilization rate of 115 mg/cm 2 -h

  11. DC plasma arc melter technology for waste vitrification

    International Nuclear Information System (INIS)

    Hamilton, R.A.; Wittle, J.K.; Trescot, J.

    1995-01-01

    This paper describes the features and benefits of a breakthrough DC Arc Melter for the permanent treatment of all types of solid wastes including nonhazardous, hazardous and radioactive. This DC Arc Furnace system, now commercially available, is the low cost permanent solution for solid waste pollution prevention and remediation. Concern over the effective disposal of wastes generated by the industrial society, worldwide, has prompted development of technologies to address the problem. For the most part these technologies have resulted in niche solutions with limited application. The only solution that has the ability to process almost all wastes, and to recover/recycle metallic and inorganic matter, is the group of technologies known as melters. Melters have distinct advantages over traditional technologies such as incineration because melters operate at higher temperatures, are relatively unaffected by changes in the waste stream, produce a vitrified stable product, and have the capability to recover/recycle slag, metals and gas. The system, DC Plasma Arc Melter, has the lowest capital, maintenance and operating cost of any melter technology because of its patented DC Plasma Arc with graphite electrode. DC Plasma Arc Melter systems are commercially available in sizes from 50 kg/batch or 250--3,000 kg/hr on a continuous feed basis. This paper examines the design and operating benefits of a DC Plasma Arc Melter System

  12. DC graphite plasma arc melter technology for waste vitrification

    International Nuclear Information System (INIS)

    Hamilton, R.A.; Wittle, J.K.; Trescot, J.; Wilver, P.

    1995-01-01

    This paper describes the features and benefits of a DC Arc Melter for the permanent treatment of all types of solid wastes including nonhazardous, hazardous and radioactive. This DC Arc Melter system is the low cost permanent solution for solid waste pollution prevention and remediation. Concern over the effective disposal of wastes generated by our industrial society, worldwide, has prompted development of technologies to address the problem. The only solution that has the ability to process almost all wastes, and to recover/recycle metallic and inorganic matter, is the group of technologies known as melters. Melters have distinct advantages over traditional technologies such as incineration because melters; operate at higher temperatures, are relatively unaffected by changes in the waste stream, produce a vitrified stable product, reduce gaseous emissions, and have the capability to recover/recycle slag, metals and gas. The system, DC Plasma Arc Melter, has the lowest capital, maintenance and operating cost of any melter technology because of its patented DC Plasma Arc with graphite electrode. DC Plasma Arc Melter systems are available in sizes from 50 kg/batch or 250-3,000 kg/hr on a continuous basis

  13. Operating experience during high-level waste vitrification at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Valenti, P.J.; Elliott, D.I.

    1999-01-01

    This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes

  14. WEEE directive application: vitrification of e-waste

    International Nuclear Information System (INIS)

    Silva, A.C.; Pidone, L.; Mello-Castanho, S.R.H.

    2011-01-01

    The aim of the WEEE directive is the electrical and electronic equipment waste (e-waste) mitigation and reduce life-cycle-end environmental impact of the this kind of equipment. Currently applied to electrical and electronic equipment manufacturers and distributors in European Union, it's Pigovian principle based, where the person that place potentially polluting products at the environment, is also legal and financial responsible for the resultant life-cycle-end pollution. This policy is a worldwide trend, and implies a growing demand in the search for new proper waste disposal technological solutions. In this context of broad spectrum, are also entered the medical equipment that make use of electronic devices. This study aims to obtain glasses which can be used as a product of commercial value containing up to 40% by mass of hospital apparatus e-waste. The residues were previously crushed and calcined. Copper and iron aventurine glass were prepared through the modification of the basic composition of soda-lime-borosilicate glass. After the melting temperature of 1300°C, the glasses were characterized by FT-IR and XRD. The chemical stability was evaluated by hydrolytic attack test. The glasses showed a high chemistry and environmental stability, like commercial soda-lime glass, in addition the obtained material present attractive appearance, allowing its use in decoration or architecture products. (author)

  15. Interim data quality objectives for waste pretreatment and vitrification. Revision 1

    International Nuclear Information System (INIS)

    Kupfer, M.J.; Conner, J.M.; Kirkbride, R.A.; Mobley, J.R.

    1994-01-01

    The Tank Waste Remediation System (TWRS) is responsible for storing, processing, and immobilizing the Hanford Site tank wastes. Characterization information on the tank wastes is needed so that safety concerns can be addressed, and retrieval, pretreatment, and immobilization processes can be designed, permitted, and implemented. This document describes the near-term tank waste sampling and characterization needs of the Pretreatment, High-Level Waste (HLW) Disposal, and Low-Level Waste (LLW) Disposal Programs to support the TWRS disposal mission. The final DQO (Data Quality Objective) will define specific waste tanks to be sampled, sample timing requirements, an appropriate analytical scheme, and a list of required analytes. This interim DQO, however, focuses primarily on the required analytes since the tanks to be sampled in FY 1994 and early FY 1995 are being driven most heavily by other considerations, particularly safety. The major objective of this Interim DQO is to provide guidance for tank waste characterization requirements for samples taken before completion of the final DQO. The characterization data needs defined herein will support the final DQO to help perform the following: Support the TWRS technical strategy by identification of the chemical and physical composition of the waste in the tanks and Guide development efforts to define waste pretreatment processes, which will in turn define HLW and LLW feed to vitrification processes

  16. Mixed Waste Working Group report

    International Nuclear Information System (INIS)

    1993-01-01

    The treatment of mixed waste remains one of this country's most vexing environmental problems. Mixed waste is the combination of radioactive waste and hazardous waste, as defined by the Resource Conservation and Recovery Act (RCRA). The Department of Energy (DOE), as the country's largest mixed waste generator, responsible for 95 percent of the Nation's mixed waste volume, is now required to address a strict set of milestones under the Federal Facility Compliance Act of 1992. DOE's earlier failure to adequately address the storage and treatment issues associated with mixed waste has led to a significant backlog of temporarily stored waste, significant quantities of buried waste, limited permanent disposal options, and inadequate treatment solutions. Between May and November of 1993, the Mixed Waste Working Group brought together stakeholders from around the Nation. Scientists, citizens, entrepreneurs, and bureaucrats convened in a series of forums to chart a course for accelerated testing of innovative mixed waste technologies. For the first time, a wide range of stakeholders were asked to examine new technologies that, if given the chance to be tested and evaluated, offer the prospect for better, safer, cheaper, and faster solutions to the mixed waste problem. In a matter of months, the Working Group has managed to bridge a gap between science and perception, engineer and citizen, and has developed a shared program for testing new technologies

  17. Remote maintenance techniques in the furnace cell of a high level waste vitrification plant

    International Nuclear Information System (INIS)

    Selig, M.

    1983-01-01

    Remote controlled maintenance and changing techniques for the furnace of a vitrification plant for radioactive waste was developed and tested on a 1:1 model. The model was fitted out with imitation main components, remote control equipment, lead-ins and the complete tubing so that the trials could be carried out in a manner replicating as closely as possible the situation found under operating conditions. The development of remote-handled tube cable connectors, tube cable jumpers and plugs and sockets was an important aspect of the developmental programme. (orig.) [de

  18. Hanford Waste Vitrification Plant technical background document for best available radionuclide control technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, A.B.; Skone, S.S.; Rodenhizer, D.G.; Marusich, M.V. (Ebasco Services, Inc., Bellevue, WA (USA))

    1990-10-01

    This report provides the background documentation to support applications for approval to construct and operate new radionuclide emission sources at the Hanford Waste Vitrification Plant (HWVP) near Richland, Washington. The HWVP is required to obtain permits under federal and state statutes for atmospheric discharges of radionuclides. Since these permits must be issued prior to construction of the facility, draft permit applications are being prepared, as well as documentation to support these permits. This report addresses the applicable requirements and demonstrates that the preferred design meets energy, environmental, and economic criteria for Best Available Radionuclide Control Technology (BARCT) at HWVP. 22 refs., 11 figs., 25 tabs.

  19. Hanford Waste Vitrification Plant technical background document for best available radionuclide control technology demonstration

    International Nuclear Information System (INIS)

    Carpenter, A.B.; Skone, S.S.; Rodenhizer, D.G.; Marusich, M.V.

    1990-10-01

    This report provides the background documentation to support applications for approval to construct and operate new radionuclide emission sources at the Hanford Waste Vitrification Plant (HWVP) near Richland, Washington. The HWVP is required to obtain permits under federal and state statutes for atmospheric discharges of radionuclides. Since these permits must be issued prior to construction of the facility, draft permit applications are being prepared, as well as documentation to support these permits. This report addresses the applicable requirements and demonstrates that the preferred design meets energy, environmental, and economic criteria for Best Available Radionuclide Control Technology (BARCT) at HWVP. 22 refs., 11 figs., 25 tabs

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

    International Nuclear Information System (INIS)

    Callow, R.A.; Weidner, J.R.; Thompson, L.E.

    1991-02-01

    This report describes the two in situ vitrification field tests conducted in June and July 1990 at 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 assess 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, indicating the process is a feasible technology for application to buried waste

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

    International Nuclear Information System (INIS)

    Callow, R.A.; Weidner, J.R.; Thompson, L.E.

    1991-01-01

    This report describes the two in situ vitrification field tests conducted in July and July 1990 at 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 assess 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 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, indicating the process is a feasible technology for application to buried waste. 8 refs., 91 figs., 13 tabs

  2. The effect of vitrification technology on waste loading

    International Nuclear Information System (INIS)

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

    1994-08-01

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

  3. Vitrification of low-level radioactive waste in a slagging combustor

    International Nuclear Information System (INIS)

    Holmes, M.J.; Downs, W.; Higley, B.A.

    1995-07-01

    The suitability of a Babcock ampersand Wilcox cyclone furnace to vitrify a low-level radioactive liquid waste was evaluated. The feed stream contained a mixture of simulated radioactive liquid waste and glass formers. The U.S. Department of Energy is testing technologies to vitrify over 60,000,000 gallons of this waste at the Hanford site. The tests reported here demonstrated the technical feasibility of Babcock ampersand Wilcox's cyclone vitrification technology to produce a glass for near surface disposal. Glass was produced over a period of 24-hours at a rate of 100 to 150 lb/hr. Based on glass analyses performed by an independent laboratory, all of the glass samples had leachabilities at least as low as those of the laboratory glass that the recipe was based upon. This paper presents the results of this demonstration, and includes descriptions of feed preparation, glass properties, system operation, and flue gas composition. The paper also provides discussions on key technical issues required to match cyclone furnace vitrification technology to this U.S. Department of Energy Hanford site application

  4. Treatment of mixed wastes by thermal plasma discharges

    International Nuclear Information System (INIS)

    Diaz A, L.V.

    2007-01-01

    The present study has as purpose to apply the technology of thermal plasma in the destruction of certain type of waste generated in the ININ. As first instance, origin, classification and disposition of the radioactive waste generated in the ININ is identified. Once identified the waste, the waste to treat is determined based on: the easiness of treating him with plasma, classification and importance. Later on, a substance or compound settles down (sample model) that serves as indicative of the waste for its physical-chemical characteristics, this is made because in the Thermal Plasma Applications Laboratory is not had the license to work with radioactive material. The sample model and the material to form the vitreous matrix are characterized before and after the treatment in order to evaluating their degradation and vitrification. During the treatment by means of the thermal plasma, the appropriate conditions are determined for the degradation and vitrification of the waste. Also, it is carried out an energy balance in the system to know the capacity to fuse the material depending the transfer of existent heat between the plasma and the material to treat. Obtaining favorable results, it thought about to climb in the project and by this way to help to solve one of the environmental problems in Mexico, as they are it the mixed wastes. (Author)

  5. Wastes vitrification by plasma torch: study of a glass formulation compatible with a wide range of B wastes

    International Nuclear Information System (INIS)

    Poitou, S.; Richaud, D.; Fiquet, O.; Gramondi, P.; Massit, H.

    2001-01-01

    Within the context of radioactive waste management, CEA has equipped itself with a 'PLASMARC' device. The central element of this device is a plasma torch treatment furnace. It has been implemented and validated for the vitrification of low level radioactive wastes. Meanwhile, the plasma torch presents potentially interests for immobilizing under an inert form in vitreous matrices, B wastes which are generally divided and of complex chemical composition. The application of this process to this type of wastes has been studied here. The obtained results show that with the plasma torch it is possible to make glasses with a high amount of silicon and aluminium oxide and which are adapted to the treatment / packaging of the B wastes. (O.M.)

  6. Feasibility study for the processing of Hanford Site cesium and strontium isotopic sources in the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Anantatmula, R.P.; Watrous, R.A.; Nelson, J.L.; Perez, J.M.; Peters, R.D.; Peterson, M.E.

    1991-09-01

    The final environmental impact statement for the disposal of defense-related wastes at the Hanford Site (Final Environmental Impact Statement: Disposal of Hanford Defense High-Level, Transuranic and Tank Wastes [HDW-EIS] [DOE 1987]) states that the preferred alternative for disposal of cesium and strontium wastes at the Hanford Site will be to package and ship these wastes to the commercial high-level waste repository. The Record of Decision for this EIS states that before shipment to a geologic repository, these wastes will be packaged in accordance with repository waste acceptance criteria. However, the high cost per canister for repository disposal and uncertainty about the acceptability of overpacked capsules by the repository suggest that additional alternative means of disposal be considered. Vitrification of the cesium and strontium salts in the Hanford Waste Vitrification Plant (HWVP) has been identified as a possible alternative to overpacking. Subsequently, Westinghouse Hanford Company's (Westinghouse Hanford) Projects Technical Support Office undertook a feasibility study to determine if any significant technical issues preclude the vitrification of the cesium and strontium salts. Based on the information presented in this report, it is considered technically feasible to blend the cesium chloride and strontium fluoride salts with neutralized current acid waste (NCAW) and/or complexant concentrate (CC) waste feedstreams, or to blend the salts with fresh frit and process the waste through the HWVP

  7. Laboratory stabilization/solidification of surrogate and actual mixed-waste sludge in glass and grout

    International Nuclear Information System (INIS)

    Spence, R.D.; Gilliam, T.M.; Mattus, C.H.; Mattus, A.J.

    1998-01-01

    Grouting and vitrification are currently the most likely stabilization/solidification technologies for mixed wastes. Grouting has been used to stabilize and solidify hazardous and low-level waste for decades. Vitrification has long been developed as a high-level-waste alternative and has been under development recently as an alternative treatment technology for low-level mixed waste. Laboratory testing has been performed to develop grout and vitrification formulas for mixed-waste sludges currently stored in underground tanks at Oak Ridge National Laboratory (ORNL) and to compare these waste forms. Envelopes, or operating windows, for both grout and soda-lime-silica glass formulations for a surrogate sludge were developed. One formulation within each envelope was selected for testing the sensitivity of performance to variations (±10 wt%) in the waste form composition and variations in the surrogate sludge composition over the range previously characterized in the sludges. In addition, one sludge sample of an actual mixed-waste tank was obtained, a surrogate was developed for this sludge sample, and grout and glass samples were prepared and tested in the laboratory using both surrogate and the actual sludge. The sensitivity testing of a surrogate tank sludge in selected glass and grout formulations is discussed in this paper, along with the hot-cell testing of an actual tank sludge sample

  8. Application of stochastic dynamic simulation to waste form qualification for the HWVP vitrification process

    International Nuclear Information System (INIS)

    Kuhn, W.L.; Westsik, J.H. Jr.

    1989-01-01

    Processing steps during the conversion of high-level nuclear waste into borosilicate glass in the Hanford Waste Vitrification Plant are being simulated on a computer by addressing transient mass balances. The results are being used to address the US Department of Energy's Waste Form Qualification requirements. The simulated addresses discontinuous (batch) operations and perturbations in the transient behavior of the process caused by errors in measurements and control actions. A collection of tests, based on process measurements, is continually checked and used to halt the simulated process when specified conditions are met. An associated set of control actions is then implemented in the simulation. The results for an example simulation are shown. 8 refs

  9. Stabilization of low-level mixed waste in chemically bonded phosphate ceramics

    International Nuclear Information System (INIS)

    Wagh, A.S.; Singh, D.; Sarkar, A.V.

    1994-06-01

    Mixed waste streams, which contain both chemical and radioactive wastes, are one of the important categories of DOE waste streams needing stabilization for final disposal. Recent studies have shown that chemically bonded phosphate ceramics may have the potential for stabilizing these waste streams, particularly those containing volatiles and pyrophorics. Such waste streams cannot be stabilized by conventional thermal treatment methods such as vitrification. Phosphate ceramics may be fabricated at room temperature into durable, hard and dense materials. For this reason room-temperature-setting phosphate ceramic waste forms are being developed to stabilize these to ''problem waste streams.''

  10. Design, operation, and evaluation of the transportable vitrification system

    International Nuclear Information System (INIS)

    Zamecnik, J.R.; Young, S.R.; Hansen, E.K.; Whitehouse, J.C.

    1997-01-01

    The Transportable Vitrification System (TVS) is a transportable melter system designed to demonstrate the treatment of low-level and mixed hazardous and radioactive wastes such as wastewater treatment sludges, contaminated soils and incinerator ash. The TVS is a large-scale, fully integrated vitrification system consisting of melter feed preparation, melter, offgas, service, and control modules. The TVS was tested with surrogate waste at the Clemson University Environmental Systems Engineering Department's (ESED) DOE/Industry Center for Vitrification Research prior to being shipped to the DOE Oak Ridge Reservation (ORR) K-25 site for treatment of mixed waste. This testing, along with additional testing at ORR, proved that the TVS would be able to successfully treat mixed waste. These surrogate tests consistently produced glass that met the EPA Toxicity Characteristic Leaching Procedure (TCLP). Performance of the system resulted in acceptable emissions of regulated metals from the offgas system. The TVS is scheduled to begin mixed waste operations at ORR in June 1997

  11. Corrosion study for a radioactive waste vitrification facility

    International Nuclear Information System (INIS)

    Imrich, K.J.; Jenkins, C.F.

    1993-01-01

    A corrosion monitoring program was setup in a scale demonstration melter system to evaluate the performance of materials selected for use in the Defense Waste Processing Facility (DWPF) at the DOE's Savannah River Site. The system is a 1/10 scale prototypic version of the DWPF. In DWPF, high activity radioactive waste will be vitrified and encapsulated for long term storage. During this study twenty-six different alloys, including DWPF reference materials of construction and alternate higher alloy materials, were subjected to process conditions and environments characteristic of the DWPF except for radioactivity. The materials were exposed to low pH, elevated temperature (to 1200 degree C) environments containing abrasive slurries, molten glass, mercury, halides and sulfides. General corrosion rates, pitting susceptibility and stress corrosion cracking of the materials were investigated. Extensive data were obtained for many of the reference materials. Performance in the Feed Preparation System was very good, whereas coupons from the Quencher Inlet region of the Melter Off-Gas System experienced localized attack

  12. Investigation of microwave melting and vitrification of nuclear wastes

    International Nuclear Information System (INIS)

    Silve, J.M.

    1996-10-01

    This work deals with the treatment process of long-lived alpha-emitting wastes, developed by the CEA, and consisting of incinerating these wastes and of vitrifying the obtained ashes by micro-waves. In this study is at first presented the materials to characterize (incineration ashes, glasses) followed by measurements of their electric permittivity. A description of the melting furnace DEMW (Densification by Micro-Waves) is given as well as its electromagnetic characterization. The inductive heating of the molten glass by micro-waves has been revealed in this part. The modeling of the power transfer to the molten glass is then presented. It is based on a sectoring of the furnace in terms of the natural weakening of each of the three areas. With this model can be obtained a general formula of the power dissipated in the molten glass in terms of the attenuation and of the position of the short-circuit piston. The fourth part relates the melt tests of different glasses and is based on the technological improvements carried out in the furnace and coming from modeling. A correlation between the modeling and the tests results is proposed. A brief characterization of the obtained material allows to verify that its qualities are acceptable particularly for leaching. (O.M.)

  13. Evaluation of alternative chemical additives for high-level waste vitrification feed preparation processing

    International Nuclear Information System (INIS)

    Seymour, R.G.

    1995-01-01

    During the development of the feed processing flowsheet for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), research had shown that use of formic acid (HCOOH) could accomplish several processing objectives with one chemical addition. These objectives included the decomposition of tetraphenylborate, chemical reduction of mercury, production of acceptable rheological properties in the feed slurry, and controlling the oxidation state of the glass melt pool. However, the DEPF research had not shown that some vitrification slurry feeds had a tendency to evolve hydrogen (H 2 ) and ammonia (NH 3 ) as the result of catalytic decomposition of CHOOH with noble metals (rhodium, ruthenium, palladium) in the feed. Testing conducted at Pacific Northwest Laboratory and later at the Savannah River Technical Center showed that the H 2 and NH 3 could evolve at appreciable rates and quantities. The explosive nature of H 2 and NH 3 (as ammonium nitrate) warranted significant mitigation control and redesign of both facilities. At the time the explosive gas evolution was discovered, the DWPF was already under construction and an immediate hardware fix in tandem with flowsheet changes was necessary. However, the Hanford Waste Vitrification Plant (HWVP) was in the design phase and could afford to take time to investigate flowsheet manipulations that could solve the problem, rather than a hardware fix. Thus, the HWVP began to investigate alternatives to using HCOOH in the vitrification process. This document describes the selection, evaluation criteria, and strategy used to evaluate the performance of the alternative chemical additives to CHOOH. The status of the evaluation is also discussed

  14. Vitrification and Testing of Hanford Pretreated Low Activity Waste

    International Nuclear Information System (INIS)

    Smith, Gary Lynn L.; Smith, Harry D.; Schweiger, Michael; Piepel, Gregory F.; Smith, Gary L.; Sundaram, S.K.; Spearing, Dane R.

    2002-01-01

    Actual pretreated LAW samples were vitrified to demonstrate the RPP-WTP projects ability to satisfy the LAW product ORP Phase B-1 contract requirements concerning, chemical and radionuclide reporting, waste loading, identification and quantification of crystalline and non-crystalline phases, and waste form leachability. Chemical compositions of two LAW glasses (i.e. elements (excluding oxygen) present in concentrations greater than 0.5 percent by weight) were measured using KOH and Na2O2 fusion preparation procedures. The measured wt% sodium oxide content for the AW-101 and AN-107 glasses are 17.7 and 18.3 respectively; however, it is argued herein that process knowledge, i.e. the target sodium oxide content, is better than the analytical measurement. Therefore for both LAW glasses the target oxide loading for sodium of 20 wt% is accepted. At these levels the glass meets or exceeds both the RPP-WTP glass specification and the DOE ORG contract requirement for waste sodium loading. The concentrations of 137Cs, 90Sr, 99Tc and transuranic (TRU) radionuclides for AW-101 and AN-107 are: (1) 0.231 and 0.292 Ci/m3, 0.435 and 0.005 Ci/m3, 0.019 and 0.129 Ci/m3, andlt; 0.16 andlt; 2.6 nCi/g, respectively. The ORP contract criteria for 137Cs, 90Sr and TRU (shall be less than 3 Ci/m3, 20 Ci/m3, and 100 nCi/g, respectively) are met in both glasses. The ORP contract criteria for 99Tc (shall be less than 0.1 Ci/m3) is met explicitly by AW-101 and will be met for the AN-107 glass by averaging its 99Tc content over the previous LAW glasses produced to meet the contract. After canister centerline cooling, no crystals were observed in the AW-101 and AN-107 glasses by XRD, optical examination and SEM analysis. The normalized PCT release rates of sodium, silicon, and boron at both 40 and 90 C from the AW-101 and AN-107 glasses are less than 2.0 g/m2 the ORP contract criteria

  15. Slurry sampling in a radioactive waste vitrification facility

    International Nuclear Information System (INIS)

    Steimke, J.L.

    1995-01-01

    The next total-system performance-assessment (TSPA) analyses are designed to aid DOE in performing an ''investment analysis'' for Yucca Mountain. This TSPA must try to bound the uncertainties for several issues that will contribute to the decision whether the US should proceed with the development of a nuclear-waste repository at Yucca Mountain. Because site-characterization experiments and data collection will continue for the foreseeable future, the next TSPA (called TSPA-IA) will again only be able to use partially developed models and partial data sets. In contrast to previous analyses however, TSPA-IA must address more specific questions to be of assistance to the investment-analysis deliberations

  16. AECL's mixed waste management program

    International Nuclear Information System (INIS)

    Peori, R.; Hulley, V.

    2006-01-01

    Every nuclear facility has it, they wish that they didn't but they have generated and do possess m ixed waste , and until now there has been no permanent disposition option; it has been for the most been simply maintained in interim storage. The nuclear industry has been responsibly developing permanent solutions for solid radioactive waste for over fifty years and for non-radioactive, chemically hazardous waste, for the last twenty years. Mixed waste (radioactive and chemically hazardous waste) however, because of its special, duo-hazard nature, has been a continuing challenge. The Hazardous Waste and Segregation Program (HW and SP) at AECL's CRL has, over the past ten years, been developing solutions to deal with their own in-house mixed waste and, as a result, have developed solutions that they would like to share with other generators within the nuclear industry. The main aim of this paper is to document and describe the early development of the solutions for both aqueous and organic liquid wastes and to advertise to other generators of this waste type how these solutions can be implemented to solve their mixed waste problems. Atomic Energy of Canada Limited (AECL) and in particular, CRL has been satisfactorily disposing of mixed waste for the last seven years. CRL has developed a program that not only disposes of mixed waste, but offers a full service mixed waste management program to customers within Canada (that could eventually include U.S. sites as well) that has developed the experience and expertise to evaluate and optimize current practices, dispose of legacy inventories, and set up an efficient segregation system to reduce and effectively manage, both the volumes and expense of, the ongoing generation of mixed waste for all generators of mixed waste. (author)

  17. Properties of the platinoid fission products during vitrification of high-level radioactive waste

    Science.gov (United States)

    Gong, W.; Lutze, W.; Perez-Cardenas, F.; Matlack, K. S.; Pegg, I. L.

    2006-05-01

    Platinoid fission products present in high-level nuclear wastes present particular challenges to their treatment by vitrification. The platinoid metals Ru, Rh, Pd, and their compounds are sparingly soluble in borosilicate glass melts. During glass melting under oxidizing conditions, the platinoids form small crystals of highly dense solid intermetallic phases and oxides. Under reducing conditions, the platinoids form only intermetallic phases. A fraction of these crystals settles to the bottom of the melting furnace, forming an immobile sludge. The fraction settling reported in the literature is highly variable. In the present work, the fraction settling was found to be >90% under reducing conditions but only 10 to 20% under oxidizing conditions. The thickness of the sludge layer depends on the volume fraction of platinoid crystals in the sludge, which is poorly known (typically ~0.06 under oxidizing conditions). Since the electrical conductivity of the sludge can be >10X that of the melt, in joule-heated melters the presence of such a layer can lead to diversion of the electric current, thereby compromising melter operability. The time to failure by this mechanism is clearly of practical importance. A variety of data are required in order to estimate the time to failure due to this mechanism and such data must be obtained under conditions representative of those in a full-size melting furnace. We have acquired such data using a melting furnace installed in our laboratory. This furnace is a one-third scale prototype of the system to be used for the vitrification of defense HLW at Hanford, WA. In the present work, simulated Hanford HLW material was combined with glass formers to produce a melter feed slurry that was then spiked with the platinoids. Over one thousand chemical and optical analyses were performed on hundreds of samples taken from the feed, various locations inside the furnace, the glass melt during pouring, the solid glass, and various locations along

  18. Sodalite as a vehicle to increase Re retention in waste glass simulant during vitrification

    Energy Technology Data Exchange (ETDEWEB)

    Luksic, Steven A., E-mail: steven.luksic@pnnl.gov; Riley, Brian J.; Parker, Kent E.; Hrma, Pavel

    2016-10-15

    Technetium (Tc) retention during Hanford waste vitrification can be increased if the volatility can be controlled. Incorporating Tc into a thermally stable mineral phase, such as sodalite, is one way to achieve increased retention. Here, rhenium (Re)-bearing sodalite was tested as a vehicle to transport perrhenate (ReO{sub 4}{sup −}), a nonradioactive surrogate for pertechnetate (TcO{sub 4}{sup −}), into high-level (HLW) and low-activity waste (LAW) glass simulants. After melting HLW and LAW simulant feeds, the retention of Re in the glass was measured and compared with the Re retention in glass prepared from a feed containing Re{sub 2}O{sub 7}. Phase analysis of sodalite in both these glasses across a profile of temperatures describes the durability of Re-sodalite during the feed-to-glass transition. The use of Re sodalite improved the Re retention by 21% for HLW glass and 85% for LAW glass, demonstrating the potential improvement in Tc-retention if TcO{sub 4}{sup −} were to be encapsulated in a Tc-sodalite prior to vitrification. - Highlights: • Re retention is improved by incorporation into sodalite structure. • LAW-type glass shows lower retention but larger improvement with Re-sodalite. • Sodalite is stable to higher temperatures in high-alumina glass melts.

  19. Vitrification treatment options for disposal of greater-than-Class-C low-level waste in a deep geologic repository

    International Nuclear Information System (INIS)

    Fullmer, K.S.; Fish, L.W.; Fischer, D.K.

    1994-11-01

    The Department of Energy (DOE), in keeping with their responsibility under Public Law 99-240, the Low-Level Radioactive Waste Policy Amendments Act of 1985, is investigating several disposal options for greater-than-Class C low-level waste (GTCC LLW), including emplacement in a deep geologic repository. At the present time vitrification, namely borosilicate glass, is the standard waste form assumed for high-level waste accepted into the Civilian Radioactive Waste Management System. This report supports DOE's investigation of the deep geologic disposal option by comparing the vitrification treatments that are able to convert those GTCC LLWs that are inherently migratory into stable waste forms acceptable for disposal in a deep geologic repository. Eight vitrification treatments that utilize glass, glass ceramic, or basalt waste form matrices are identified. Six of these are discussed in detail, stating the advantages and limitations of each relative to their ability to immobilize GTCC LLW. The report concludes that the waste form most likely to provide the best composite of performance characteristics for GTCC process waste is Iron Enriched Basalt 4 (IEB4)

  20. French industrial plant AVM for continuous vitrification of high level radioactive wastes

    International Nuclear Information System (INIS)

    Bonniaud, Roger; Sombret, Claude; Barbe, Alain

    1975-01-01

    The A.V.M. plant is a continuous process plant now under construction at Marcoule and intended for vitrifying the whole of fission product solutions from the C.E.A. (Commissariat a l'Energie Atomique) - Marcoule reprocessing plant. The outset of the construction took place in the second 1974 half year; the first radioactive run is scheduled in July 1977. The two steps of the process are shown: first a continuous calcination then a continuous glass making from the calcined product and suitable additives. The plant consists in two parts: vitrification and storage facilities. Some wastes will be continuously produced day after day due to gas clean up and worn out materials. Characteristics of the solutions processed, calcined products, glass composition, and expected liquid wastes are given in tables [fr

  1. Low-Level Waste Vitrification Plant Project contracting strategy decision analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Felise, P.; Phillips, J.D.

    1994-10-17

    Ten basic contracting strategies were developed after a review of past strategies that had been used at the Hanford Site, other US Department of Energy (DOE) sites, other US government agencies, and in the private sector. As applicable to the Low-Level Waste Vitrification Plant (LLWVP) Project, each strategy was described and depicted in a schedule format to assess compatibility with the Hanford Federal Facility Agreement and Consent Order, al so known as the Tri-Party Agreement (Ecology et al. 1994) milestones, key decision points, and other project requirements. The-pro and con aspects of each strategy also were tabulated. Using this information as a basis, the LLWVP Project team members, along with representatives of Tank Waste Remediation System (TWRS) Engineering, TWRS Programs, and Procurement Materials Management, formed a Westinghouse Hanford Company (WHC) evaluation team to select the best strategy. Kepner-Tregoe decision analysis techniques were used in facilitated meetings to arrive at the best balanced choice.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

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

  3. Human Factors engineering criteria and design for the Hanford Waste Vitrification Plant preliminary safety analysis report

    International Nuclear Information System (INIS)

    Wise, J.A.; Schur, A.; Stitzel, J.C.L.

    1993-09-01

    This report provides a rationale and systematic methodology for bringing Human Factors into the safety design and operations of the Hanford Waste Vitrification Plant (HWVP). Human Factors focuses on how people perform work with tools and machine systems in designed settings. When the design of machine systems and settings take into account the capabilities and limitations of the individuals who use them, human performance can be enhanced while protecting against susceptibility to human error. The inclusion of Human Factors in the safety design of the HWVP is an essential ingredient to safe operation of the facility. The HWVP is a new construction, nonreactor nuclear facility designed to process radioactive wastes held in underground storage tanks into glass logs for permanent disposal. Its design and mission offer new opposites for implementing Human Factors while requiring some means for ensuring that the Human Factors assessments are sound, comprehensive, and appropriately directed

  4. Low-Level Waste Vitrification Plant Project contracting strategy decision analysis report

    International Nuclear Information System (INIS)

    Felise, P.; Phillips, J.D.

    1994-01-01

    Ten basic contracting strategies were developed after a review of past strategies that had been used at the Hanford Site, other US Department of Energy (DOE) sites, other US government agencies, and in the private sector. As applicable to the Low-Level Waste Vitrification Plant (LLWVP) Project, each strategy was described and depicted in a schedule format to assess compatibility with the Hanford Federal Facility Agreement and Consent Order, al so known as the Tri-Party Agreement (Ecology et al. 1994) milestones, key decision points, and other project requirements. The-pro and con aspects of each strategy also were tabulated. Using this information as a basis, the LLWVP Project team members, along with representatives of Tank Waste Remediation System (TWRS) Engineering, TWRS Programs, and Procurement Materials Management, formed a Westinghouse Hanford Company (WHC) evaluation team to select the best strategy. Kepner-Tregoe decision analysis techniques were used in facilitated meetings to arrive at the best balanced choice

  5. Cold-cap reactions in vitrification of nuclear waste glass: experiments and modeling

    International Nuclear Information System (INIS)

    Chun, Jaehun; Pierce, David A.; Pokorny, Richard; Hrma, Pavel R.

    2013-01-01

    Cold-cap reactions are multiple overlapping reactions that occur in the waste-glass melter during the vitrification process when the melter feed is being converted to molten glass. In this study, we used differential scanning calorimetry (DSC) to investigate cold-cap reactions in a high-alumina high-level waste melter feed. To separate the reaction heat from both sensible heat and experimental instability, we employed the run/rerun method, which enabled us to define the degree of conversion based on the reaction heat and to estimate the heat capacity of the reacting feed. Assuming that the reactions are nearly independent and can be approximated by the nth order kinetics, we obtained the kinetic parameters using the Kissinger method combined with least squares analysis. The resulting mathematical simulation of the cold-cap reactions provides a key element for the development of an advanced cold-cap model

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

    International Nuclear Information System (INIS)

    Ramsey, W.G.; Gray, M.F.; Calmus, R.B.; Edge, J.A.; Garrett, B.G.

    2011-01-01

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

  7. Managing a mixed waste program

    International Nuclear Information System (INIS)

    Koch, J.D.

    1994-01-01

    IT Corporation operates an analytical laboratory in St. Louis capable of analyzing environmental samples that are contaminated with both chemical and radioactive materials. Wastes generated during these analyses are hazardous in nature; some are listed wastes others exhibit characteristic hazards. When the original samples contain significant quantities of radioactive material, the waste must be treated as a mixed waste. A plan was written to document the waste management program describing the management of hazardous, radioactive and mixed wastes. This presentation summarizes the methods employed by the St. Louis facility to reduce personnel exposures to the hazardous materials, minimize the volume of mixed waste and treat the materials prior to disposal. The procedures that are used and the effectiveness of each procedure will also be discussed. Some of the lessons that have been learned while dealing with mixed wastes will be presented as well as the solutions that were applied. This program has been effective in reducing the volume of mixed waste that is generated. The management program also serves as a method to manage the costs of the waste disposal program by effectively segregating the different wastes that are generated

  8. Mathematical Modeling of Conversion Kinetics during Vitrification of Nuclear Waste

    International Nuclear Information System (INIS)

    Pokorny, Richard; Pierce, David A.; Chun, Jae Hun; Hrma, Pavel

    2012-01-01

    The last part of the high-level waste (HLW) glass melter that has not yet been fully understood, not to mention mathematically modeled, is the cold cap. Cold cap is a layer of dry melter feed, a mixture of the HLW with glass forming and modifying additives. It floats on the pool of molten glass from which it receives the heat necessary for melting. Mathematical modeling of the cold cap solves differential equations that express the mass and energy balances for the feed-to-glass conversion within the cold cap. The feed-to-glass conversion consists of multiple chemical reactions and phase transitions. Reaction enthalpies and mass losses to gases evolved provide an important input for the cold cap modeling. In this study, we measured the kinetics of cold cap reactions using the non-isothermal thermo-gravimetric analysis (TGA) and differential scanning calorimetry (DSC). These thermoanalytical techniques show multiple overlapping peaks, necessitating the development of a deconvolution method for the determination of the kinetics of major reactions needed for cold cap modeling. Assuming that the cold cap reactions are independent, we expressed the overall rate as a sum of rates of individual reactions that we treat as Arrheniustype processes with a power-law based kinetics. Accordingly, we fitted to experimental data the following equation: dx/dT=1/Φ N Σ 1 w i A i (1-x i ) ni exp(-B i /T) (1) where x is the fraction of material reacted, T is temperature, Φ is the heating rate, wi the weight of the i th reaction (the fraction of the total mass loss caused by the i th reaction), Ai is the i th reaction pre-exponential factor, B i is the i th reaction activation energy, and n i is the i th reaction (apparent) reaction order. Because HLW melter feeds contain a large number of constituents, such as oxides, acids, hydroxides, oxyhydrates, and ionic salts, the number of cold cap reactions is very large indeed. For example, hydroxides, oxyhydrates, boric acid, and various

  9. Evaporation, Rheology, And Vitrification Of A Pretreated Radioactive Hanford Tank 241-AN-104 Sample Mixed With Simulated Law SBS Recycle

    International Nuclear Information System (INIS)

    MARK, CROWDER

    2004-01-01

    This study involved evaporation of the radioactive low activity waste AN-104 pretreated waste and a simulant LAW submerged bed scrubber recycle from Duratek blended at two different volume ratios. The AN104SBS35651 pretreated wastes were then blended with glass former chemicals, GFCs, and a single blend vitrified. The chemical and physical properties, during all phases of blending were characterized per Table 1-1. The AN-104 radioactive waste used for this study was initially characterized at SRNL, Hay 2003, followed by filtration to remove entrained solids, Poirier 2003, and put through ion exchange for cesium removal, Adu-Wusu 2003,. All the test objectives in Table 1-1 are from section 3 of the Test Specification, Sidibe 2003. The test exception listed in Table 1-2 deleted the vitrification objectives for vitrification product analyses and product testing of the AN-104 glass. Therefore, Table 1-1 does not list any vitrification product analyses and product testing objectives

  10. Vitrification publication bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Schmieman, E.; Johns, W.E.

    1996-02-01

    This document was compiled by a group of about 12 graduate students in the Department of Mechanical Engineering and Material Science at Washington State University and was funded by the U.S. Department of Energy. The literature search resulting in the compilation of this bibliography was designed to be an exhaustive search for research and development work involving the vitrification of mixed wastes, published by domestic and foreign researchers, primarily during 1989-1994. The search techniques were dominated by electronic methods and this bibliography is also available in electronic format, Windows Reference Manager.

  11. Vitrification publication bibliography

    International Nuclear Information System (INIS)

    Schmieman, E.; Johns, W.E.

    1996-02-01

    This document was compiled by a group of about 12 graduate students in the Department of Mechanical Engineering and Material Science at Washington State University and was funded by the U.S. Department of Energy. The literature search resulting in the compilation of this bibliography was designed to be an exhaustive search for research and development work involving the vitrification of mixed wastes, published by domestic and foreign researchers, primarily during 1989-1994. The search techniques were dominated by electronic methods and this bibliography is also available in electronic format, Windows Reference Manager

  12. US DOE Initiated Performance Enhancements to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-activity Waste Vitrification (LAW) System

    International Nuclear Information System (INIS)

    Hamel, William F.; Gerdes, Kurt D.; Holton, Langdon K.; Pegg, Ian L.; Bowen, Brad W.

    2006-01-01

    The U.S Department of Energy Office of River Protection (DOE-ORP) is constructing a Waste Treatment and Immobilization Plant (WTP) for the treatment and vitrification of underground tank wastes stored at the Hanford Site in Washington State. The WTP comprises four major facilities: a pretreatment facility to separate the tank waste into high level waste (HLW) and low-activity waste (LAW) process streams, a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction, and an analytical laboratory to support the operations of all four treatment facilities. DOE has established strategic objectives to optimize the performance of the WTP facilities and the LAW and HLW waste forms to reduce the overall schedule and cost for treatment and vitrification of the Hanford tank wastes. This strategy has been implemented by establishing performance expectations in the WTP contract for the facilities and waste forms. In addition, DOE, as owner-operator of the WTP facilities, continues to evaluate (1) the design, to determine the potential for performance above the requirements specified in the WTP contract; and (2) improvements in production of the LAW and HLW waste forms. This paper reports recent progress directed at improving production of the LAW waste form. DOE's initial assessment, which is based on the work reported in this paper, is that the capacity of the WTP LAW vitrification facility can be increased by a factor of 2 to 4 with a combination of revised glass formulations, modest increases in melter glass operating temperatures, and a second-generation LAW melter with a larger surface area. Implementing these improvements in the LAW waste immobilization capability can benefit the LAW treatment mission by reducing both processing time and cost

  13. Hanford Waste Vitrification Plant Quality Assurance Program description for defense high-level waste form development and qualification

    International Nuclear Information System (INIS)

    Hand, R.L.

    1992-01-01

    This document describes the quality assurance (QA) program of the Hanford Waste Vitrification Plant (HWVP) Project. The purpose of the QA program is to control project activities in such a manner as to achieve the mission of the HWVP Project in a safe and reliable manner. A major aspect of the HWVP Project QA program is the control of activities that relate to high-level waste (HLW) form development and qualification. This document describes the program and planned actions the Westinghouse Hanford Company (Westinghouse Hanford) will implement to demonstrate and ensure that the HWVP Project meets the US Department of Energy (DOE) and ASME regulations. The actions for meeting the requirements of the Waste Acceptance Preliminary Specifications (WAPS) will be implemented under the HWVP product qualification program with the objective of ensuring that the HWVP and its processes comply with the WAPS established by the federal repository

  14. Evaluation of high-level waste vitrification feed preparation chemistry for an NCAW simulant, FY 1994: Alternate flowsheets (DRAFT)

    International Nuclear Information System (INIS)

    Smith, H.D.; Merz, M.D.; Wiemers, K.D.; Smith, G.L.

    1996-02-01

    High-level radioactive waste stored in tanks at the U.S. Department of Energy's (DOE's) Hanford Site will be pretreated to concentrate radioactive constituents and fed to the vitrification plant A flowsheet for feed preparation within the vitrification plant (based on the Hanford Waste Vitrification Plant (HWVP) design) called for HCOOH addition during the feed preparation step to adjust rheology and glass redox conditions. However, the potential for generating H 2 and NH 3 during treatment of high-level waste (HLW) with HCOOH was identified at Pacific Northwest Laboratory (PNL). Studies at the University of Georgia, under contract with Savannah River Technology Center (SRTC) and PNL, have verified the catalytic role of noble metals (Pd, Rh, Ru), present in the waste, in the generation of H 2 and NH 3 . Both laboratory-scale and pilot-scale studies at SRTC have documented the H 2 and NH 3 generation phenomenal Because H 2 and NH 3 may create hazardous conditions in the vessel vapor space and offgas system of a vitrification plant, reducing the H 2 generation rate and the NH 3 generation to the lowest possible levels consistent with desired melter feed characteristics is important. The Fiscal Year 1993 and 1994 studies were conducted with simulated (non-radioactive), pre-treated neutralized current acid waste (NCAW). Neutralized current acid waste is a high-level waste originating from the plutonium/uranium extraction (PUREX) plant that has been partially denitrated with sugar, neutralized with NaOH, and is presently stored in double-shell tanks. The non-radioactive simulant used for the present study includes all of the trace components found in the waste, or substitutes a chemically similar element for radioactive or very toxic species. The composition and simulant preparation steps were chosen to best simulate the chemical processing characteristics of the actual waste

  15. Secondary Waste Considerations for Vitrification of Sodium-Bearing Waste at the Idaho Nuclear Technology and Engineering Center FY-2001 Status Report

    International Nuclear Information System (INIS)

    Herbst, A.K.; Kirkham, R.J.; Losinski, S.J.

    2002-01-01

    The Idaho Nuclear Technology and Engineering Center (INTEC) is considering vitrification to process liquid sodium-bearing waste. Preliminary studies were completed to evaluate the potential secondary wastes from the melter off-gas clean up systems. Projected secondary wastes comprise acidic and caustic scrubber solutions, HEPA filters, activated carbon, and ion exchange media. Possible treatment methods, waste forms, and disposal sites are evaluated from radiological and mercury contamination estimates

  16. Secondary Waste Considerations for Vitrification of Sodium-Bearing Waste at the Idaho Nuclear Technology and Engineering Center FY-2001 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, A.K.; Kirkham, R.J.; Losinski, S.J.

    2002-09-26

    The Idaho Nuclear Technology and Engineering Center (INTEC) is considering vitrification to process liquid sodium-bearing waste. Preliminary studies were completed to evaluate the potential secondary wastes from the melter off-gas clean up systems. Projected secondary wastes comprise acidic and caustic scrubber solutions, HEPA filters, activated carbon, and ion exchange media. Possible treatment methods, waste forms, and disposal sites are evaluated from radiological and mercury contamination estimates.

  17. Secondary Waste Considerations for Vitrification of Sodium-Bearing Waste at the Idaho Nuclear Techology and Engineering Center FY-2001 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, Alan Keith; Kirkham, Robert John; Losinski, Sylvester John

    2001-09-01

    The Idaho Nuclear Technology and Engineering Center (INTEC) is considering vitrification to process liquid sodium-bearing waste. Preliminary studies were completed to evaluate the potential secondary wastes from the melter off-gas clean up systems. Projected secondary wastes comprise acidic and caustic scrubber solutions, HEPA filters, activated carbon, and ion exchange media. Possible treatment methods, waste forms, and disposal sites are evaluated from radiological and mercury contamination estimates.

  18. Impact of Alkali Source on Vitrification of SRS High Level Waste

    International Nuclear Information System (INIS)

    LAMBERT, D. P.; MILLER, D. H.; PEELER, D. K.; SMITH, M. E.; STONE, M. E.

    2005-01-01

    The Defense Waste Processing Facility (DWPF) Savannah River Site is currently immobilizing high level nuclear waste sludge by vitrification in borosilicate glass. The processing strategy involves blending a large batch of sludge into a feed tank, washing the sludge to reduce the amount of soluble species, then processing the large ''sludge batch'' through the DWPF. Each sludge batch is tested by the Savannah River National Laboratory (SRNL) using simulants and tests with samples of the radioactive waste to ''qualify'' the batch prior to processing in the DWPF. The DWPF pretreats the sludge by first acidifying the sludge with nitric and formic acid. The ratio of nitric to formic acid is adjusted as required to target a final glass composition that is slightly reducing (the target is for ∼20% of the iron to have a valence of two in the glass). The formic acid reduces the mercury in the feed to elemental mercury which is steam stripped from the feed. After a concentration step, the glass former (glass frit) is added as a 50 wt% slurry and the batch is concentrated to approximately 50 wt% solids. The feed slurry is then fed to a joule heated melter maintained at 1150 C. The glass must meet both processing (e.g., viscosity and liquidus temperature) and product performance (e.g., durability) constraints The alkali content of the final waste glass is a critical parameter that affects key glass properties (such as durability) as well as the processing characteristics of the waste sludge during the pretreatment and vitrification processes. Increasing the alkali content of the glass has been shown to improve the production rate of the DWPF, but the total alkali in the final glass is limited by constraints on glass durability and viscosity. Two sources of alkali contribute to the final alkali content of the glass: sodium salts in the waste supernate and sodium and lithium oxides in the glass frit added during pretreatment processes. Sodium salts in the waste supernate can

  19. Hanford waste vitrification plant hydrogen generation study: Preliminary evaluation of alternatives to formic acid

    International Nuclear Information System (INIS)

    King, R.B.; Bhattacharyya, N.K.; Kumar, V.

    1996-02-01

    Oxalic, glyoxylic, glycolic, malonic, pyruvic, lactic, levulinic, and citric acids as well as glycine have been evaluated as possible substitutes for formic acid in the preparation of feed for the Hanford waste vitrification plant using a non-radioactive feed stimulant UGA-12M1 containing substantial amounts of aluminum and iron oxides as well as nitrate and nitrite at 90C in the presence of hydrated rhodium trichloride. Unlike formic acid none of these carboxylic acids liberate hydrogen under these conditions and only malonic and citric acids form ammonia. Glyoxylic, glycolic, malonic, pyruvic, lactic, levulinic, and citric acids all appear to have significant reducing properties under the reaction conditions of interest as indicated by the observation of appreciable amounts of N 2 O as a reduction product of,nitrite or, less likely, nitrate at 90C. Glyoxylic, pyruvic, and malonic acids all appear to be unstable towards decarboxylation at 90C in the presence of Al(OH) 3 . Among the carboxylic acids investigated in this study the α-hydroxycarboxylic acids glycolic and lactic acids appear to be the most interesting potential substitutes for formic acid in the feed preparation for the vitrification plant because of their failure to produce hydrogen or ammonia or to undergo decarboxylation under the reaction conditions although they exhibit some reducing properties in feed stimulant experiments

  20. Mixed Waste Focus Area - Waste form initiative

    International Nuclear Information System (INIS)

    Nakaoka, R.; Waters, R.; Pohl, P.; Roach, J.

    1998-01-01

    The mission of the US Department of Energy's (DOE) Mixed Waste Focus Area (MWFA) is to provide acceptable technologies that enable implementation of mixed waste treatment systems which are developed in partnership with end-users, stakeholders, tribal governments, and regulators. To accomplish this mission, a technical baseline was established in 1996 and revised in 1997. The technical baseline forms the basis for determining which technology development activities will be supported by the MWFA. The primary attribute of the technical baseline is a set of prioritized technical deficiencies or roadblocks related to implementation of mixed waste treatment systems. The Waste Form Initiative (WFI) was established to address an identified technical deficiency related to waste form performance. The primary goal of the WFI was to ensure that the mixed low-level waste (MLLW) treatment technologies being developed, currently used, or planned for use by DOE would produce final waste forms that meet the waste acceptance criteria (WAC) of the existing and/or planned MLLW disposal facilities. The WFI was limited to an evaluation of the disposal requirements for the radioactive component of MLLW. Disposal requirements for the hazardous component are dictated by the Resource Conservation and Recovery Act (RCRA), and were not addressed. This paper summarizes the technical basis, strategy, and results of the activities performed as part of the WFI

  1. Mixed Waste Integrated Program -- Problem-oriented technology development

    International Nuclear Information System (INIS)

    Hart, P.W.; Wolf, S.W.; Berry, J.B.

    1994-01-01

    The Mixed Waste Integrated Program (MWIP) is responding to the need for DOE mixed waste treatment technologies that meet these dual regulatory requirements. MWIP is developing emerging and innovative treatment technologies to determine process feasibility. Technology demonstrations will be used to determine whether processes are superior to existing technologies in reducing risk, minimizing life-cycle cost, and improving process performance. Technology development is ongoing in technical areas required to process mixed waste: materials handling, chemical/physical treatment, waste destruction, off-gas treatment, final forms, and process monitoring/control. MWIP is currently developing a suite of technologies to process heterogeneous waste. One robust process is the fixed-hearth plasma-arc process that is being developed to treat a wide variety of contaminated materials with minimal characterization. Additional processes encompass steam reforming, including treatment of waste under the debris rule. Advanced off-gas systems are also being developed. Vitrification technologies are being demonstrated for the treatment of homogeneous wastes such as incinerator ash and sludge. An alternative to conventional evaporation for liquid removal--freeze crystallization--is being investigated. Since mercury is present in numerous waste streams, mercury removal technologies are being developed

  2. Hanford High-Level Waste Vitrification Program at the Pacific Northwest National Laboratory: technology development - annotated bibliography

    International Nuclear Information System (INIS)

    Larson, D.E.

    1996-09-01

    This report provides a collection of annotated bibliographies for documents prepared under the Hanford High-Level Waste Vitrification (Plant) Program. The bibliographies are for documents from Fiscal Year 1983 through Fiscal Year 1995, and include work conducted at or under the direction of the Pacific Northwest National Laboratory. The bibliographies included focus on the technology developed over the specified time period for vitrifying Hanford pretreated high-level waste. The following subject areas are included: General Documentation; Program Documentation; High-Level Waste Characterization; Glass Formulation and Characterization; Feed Preparation; Radioactive Feed Preparation and Glass Properties Testing; Full-Scale Feed Preparation Testing; Equipment Materials Testing; Melter Performance Assessment and Evaluations; Liquid-Fed Ceramic Melter; Cold Crucible Melter; Stirred Melter; High-Temperature Melter; Melter Off-Gas Treatment; Vitrification Waste Treatment; Process, Product Control and Modeling; Analytical; and Canister Closure, Decontamination, and Handling

  3. In situ vitrification of transuranic wastes: An updated systems evaluation and applications assessment

    Energy Technology Data Exchange (ETDEWEB)

    Buelt, J.L.; Timmerman, C.L.; Oma, K.H.; FitzPatrick, V.F.; Carter, J.G.

    1987-03-01

    In situ vitrification (ISV) is a process whereby joule heating immobilizes contaminated soil in place into a durable glass and crystalline waste form. Numerous technological advances made during the past three years in the design, fabrication, and testing of the ISV process are discussed. Performance analysis of ISV focuses on process equipment, element retention (in the vitrified soil during processing), melt geometry, depth monitors, and electrodes. The types of soil and waste processed by ISV are evaluated as process parameters. Economic data provide the production costs of the large-scale unit for radioactive and hazardous chemical wastes (wet and dry). The processing of transuranic-contaminated soils are discussed with respect to occupational and public safety. Alternative applications and operating sequences for various waste sites are identified. The technological data base warrants conducting a large-scale radioactive test at a contaminated soil site at Hanford to provide a representative waste form that can be evaluated to determine its suitability for in-place stabilization of transuranic-contaminated soils.

  4. Mixed waste characterization reference document

    International Nuclear Information System (INIS)

    1997-09-01

    Waste characterization and monitoring are major activities in the management of waste from generation through storage and treatment to disposal. Adequate waste characterization is necessary to ensure safe storage, selection of appropriate and effective treatment, and adherence to disposal standards. For some wastes characterization objectives can be difficult and costly to achieve. The purpose of this document is to evaluate costs of characterizing one such waste type, mixed (hazardous and radioactive) waste. For the purpose of this document, waste characterization includes treatment system monitoring, where monitoring is a supplement or substitute for waste characterization. This document establishes a cost baseline for mixed waste characterization and treatment system monitoring requirements from which to evaluate alternatives. The cost baseline established as part of this work includes costs for a thermal treatment technology (i.e., a rotary kiln incinerator), a nonthermal treatment process (i.e., waste sorting, macronencapsulation, and catalytic wet oxidation), and no treatment (i.e., disposal of waste at the Waste Isolation Pilot Plant (WIPP)). The analysis of improvement over the baseline includes assessment of promising areas for technology development in front-end waste characterization, process equipment, off gas controls, and monitoring. Based on this assessment, an ideal characterization and monitoring configuration is described that minimizes costs and optimizes resources required for waste characterization

  5. Feasibility Study for Preparation and Use of Glass Grains as an Alternative to Glass Nodules for Vitrification of Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Sonavane, M S; Mishra, P.K., E-mail: maheshss@barc.gov.in [Nuclear Recycle Board, Bhabha Atomic Research Centre, Mumbai (India); Mandal, S; Barik, S; Roy Chowdhury, A; Sen, R [Central Glass and Ceramic Institute, Kolkata (India)

    2012-10-15

    High level nuclear liquid waste (HLW) is immobilized using borosilicate glass matrix. Presently joule heated ceramic melter is being employed for vitrification of HLW in India. Preformed nodules of base glass are fed to melter along with liquid waste in predetermined ratio. In order to reduce the cost incurred for production of glass nodules of base glass, an alternative option of using glass grains was evaluated for its preparation and its suitability for the melter operation. (author)

  6. Feasibility Study for Preparation and Use of Glass Grains as an Alternative to Glass Nodules for Vitrification of Nuclear Waste

    International Nuclear Information System (INIS)

    Sonavane, M.S.; Mishra, P.K.; Mandal, S.; Barik, S.; Roy Chowdhury, A.; Sen, R.

    2012-01-01

    High level nuclear liquid waste (HLW) is immobilized using borosilicate glass matrix. Presently joule heated ceramic melter is being employed for vitrification of HLW in India. Preformed nodules of base glass are fed to melter along with liquid waste in predetermined ratio. In order to reduce the cost incurred for production of glass nodules of base glass, an alternative option of using glass grains was evaluated for its preparation and its suitability for the melter operation. (author)

  7. Managing mixed wastes: technical issues

    International Nuclear Information System (INIS)

    Lytle, J.E.; Eyman, L.D.; Burton, D.W.; McBrayer, J.F.

    1986-01-01

    The US Department of Energy manages wastes that are both chemically hazardous and radioactive. These mixed wastes are often unique and many have national security implications. Management practices have evolved over the more than forty years that the Department and its predecessor agencies have been managing these wastes, both in response to better understanding of the hazards involved and in response to external, regulatory influences. The Department has recently standarized its waste management practices and has initited an R and D program to address priority issues identified by its operating contractor organizations. The R and D program is guided by waste management strategy that emphasizes reduction of human exposure to hazardous wastes in the environment, reduction of the amount and toxicity of wastes generated, treatment of wastes that are generated to reduce volumes and toxicities, and identification of alternatives to land disposal of wastes that remain hazardous following maximum practicable treatment

  8. Mixed wasted integrated program: Logic diagram

    International Nuclear Information System (INIS)

    Mayberry, J.; Stelle, S.; O'Brien, M.; Rudin, M.; Ferguson, J.; McFee, J.

    1994-01-01

    The Mixed Waste Integrated Program Logic Diagram was developed to provide technical alternative for mixed wastes projects for the Office of Technology Development's Mixed Waste Integrated Program (MWIP). Technical solutions in the areas of characterization, treatment, and disposal were matched to a select number of US Department of Energy (DOE) treatability groups represented by waste streams found in the Mixed Waste Inventory Report (MWIR)

  9. Final technical report: Atmospheric emission analysis for the Hanford Waste Vitrification plant

    International Nuclear Information System (INIS)

    Andrews, G.L.; Rhoads, K.C.

    1996-03-01

    This report is an assessment of chemical and radiological effluents that are expected to be released to the atmosphere from the Hanford Waste Vitrification Plant (HWVP). The report is divided into two sections. In the first section, the impacts of carbon monoxide (CO) and nitrogen oxides as NO 2 have been estimated for areas within the Hanford Site boundary. A description of the dispersion model used to-estimate CO and NO 2 average concentrations and Hanford Site meteorological data has been included in this section. In the second section, calculations were performed to estimate the potential radiation doses to a maximally exposed off-site individual. The model used to estimate the horizontal and vertical dispersion of radionuclides is also discussed

  10. Test methods for selection of materials of construction for high-level radioactive waste vitrification. Revision

    International Nuclear Information System (INIS)

    Bickford, D.F.; Corbett, R.A.; Morrison, W.S.

    1986-01-01

    Candidate materials of construction were evaluated for a facility at the Department of Energy's Savannah River Plant to vitrify high-level radioactive waste. Limited operating experience was available under the corrosive conditions of the complex vitrification process. The objective of the testing program was to provide a high degree of assurance that equipment will meet or exceed design lifetimes. To meet this objective in reasonable time and minimum cost, a program was designed consisting of a combination of coupon immersion and electrochemical laboratory tests and pilot-scale tests. Stainless steels and nickel-based alloys were tested. Alloys that were most resistant to general and local attack contained nickel, molybdenum (>9%), and chromium (where Cr + Mo > 30%). Alloy C-276 was selected as the reference material for process equipment. Stellite 6 was selected for abrasive service in the presence of formic acid. Alloy 690 and ALLCORR were selected for specific applications

  11. Design of off-gas cleaning systems for high-level waste vitrification

    International Nuclear Information System (INIS)

    Hanson, M.S.; Kaser, J.D.

    1976-01-01

    High-level wastes are generally nitric acid solutions. Vitrification converts the nitrate salts to oxides, forming nitrogen oxides (NO/sub x/) as a by-product. These NO/sub x/ releases can be controlled by nitric acid recovery or by conversion of the NO/sub x/ to an acceptable species for release, such as N 2 O or N 2 . The off-gas system must also be capable of controlling any fission products which may be voltatilized in appreciable quantities and may be controlled in the off-gas system by absorption or adsorption. Whichever method is used, the recovered fission products must somehow be converted to a safe disposal form. Proposed off-gas systems are described, and areas requiring research and development are discussed

  12. Vitrification of Simulated Fernald K-65 Silo Waste at Low Temperature

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1999-01-01

    Vitrification is the technology that has been chosen to solidify approximately 18,000 tons of geologic mill tailings at the Fernald Environmental Management Project (FEMP) in Fernald, Ohio. The geologic mill tailings are residues from the processing of pitchlende ore during 1949-1958. These waste residues are contained in silos in Operable Unit 4 (OU4) at the FEMP facility. Operable Unit 4 is one of five operable units at the FEMP. Operable Unit 4 is one of five operable units at the FEMP. Operating Unit 4 consists of four concrete storage silos and their contents. Silos 1 and 2 contain K-65 mill tailing residues and a bentonite cap, Silo 3 contains non-radioactive metal oxides, and Silo 4 is empty

  13. In situ vitrification demonstration for the stabilization of buried wastes at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Jacobs, G.K.; Spalding, B.P.; Carter, J.G.; Koegler, S.S.

    1987-01-01

    A demonstration of In Situ Vitrification (ISV) technology for the stabilization of radioactively contaminated soil sites at the Oak Ridge National Laboratory (ORNL) was successfully completed during July 1987. This demonstration is the first application of the ISV process not performed at the Hanford Site, where the technology was developed and patented by Pacific Northwest Laboratory (PNL). The joint ORNL-PNL pilot-scale demonstration was performed on a 3/8-scale trench (2 m deep x 1 m wide x 10 m long) that was constructed to simulate a typical seepage trench used for liquid low-level radioactive waste disposal at ORNL from 1951 to 1966. In the ISV process, electrodes are inserted around a volume of contaminated soil, power is applied to the electrodes, and the entire mass is melted from the surface of the soil down through the contaminated zone, thus making a glassy-to-microcrystalline waste form that incorporates the contaminants. Gases produced during the melting are collected, treated, monitored, and released through an off-gas process trailer. In the ORNL demonstration, a 25-t mass of melted rock approximately 1.2 m thick x 2.1 m wide x 4.9 m long was formed during 110 h of operation that consumed approximately 29 MWh of power. Data obtained on the operational performance of the test and waste-form durability will be used to assess the feasibility of applying the ISV technology to an actual waste trench

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

    Energy Technology Data Exchange (ETDEWEB)

    Elmore, M R; Jensen, G A

    1991-07-01

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

  15. Mixed Waste Landfill Integrated Demonstration

    International Nuclear Information System (INIS)

    1994-02-01

    The mission of the Mixed Waste Landfill Integrated Demonstration (MWLID) is to demonstrate, in contaminated sites, new technologies for clean-up of chemical and mixed waste landfills that are representative of many sites throughout the DOE Complex and the nation. When implemented, these new technologies promise to characterize and remediate the contaminated landfill sites across the country that resulted from past waste disposal practices. Characterization and remediation technologies are aimed at making clean-up less expensive, safer, and more effective than current techniques. This will be done by emphasizing in-situ technologies. Most important, MWLID's success will be shared with other Federal, state, and local governments, and private companies that face the important task of waste site remediation. MWLID will demonstrate technologies at two existing landfills. Sandia National Laboratories' Chemical Waste Landfill received hazardous (chemical) waste from the Laboratory from 1962 to 1985, and the Mixed-Waste Landfill received hazardous and radioactive wastes (mixed wastes) over a twenty-nine year period (1959-1988) from various Sandia nuclear research programs. Both landfills are now closed. Originally, however, the sites were selected because of Albuquerque's and climate and the thick layer of alluvial deposits that overlay groundwater approximately 480 feet below the landfills. This thick layer of ''dry'' soils, gravel, and clays promised to be a natural barrier between the landfills and groundwater

  16. Test plan for glass melter system technologies for vitrification of hign-sodium content low-level radioactive liquid waste, Project No. RDD-43288

    International Nuclear Information System (INIS)

    Higley, B.A.

    1995-01-01

    This document provides a test plan for the conduct of combustion fired cyclone vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System, Low-Level Waste Vitrification Program. The vendor providing this test plan and conducting the work detailed within it is the Babcock ampersand Wilcox Company Alliance Research Center in Alliance, Ohio. This vendor is one of seven selected for glass melter testing

  17. Solidification and vitrification life-cycle economics study

    International Nuclear Information System (INIS)

    Gimpel, R.F.

    1992-01-01

    Solidification (making concrete) and vitrification (making glass) are frequently the treatment methods recommended for treating inorganic or radioactive wastes. Solidification is generally perceived as the most economical treatment method, whereas vitrification is considered (by many) as the most effective of all treatment methods. Unfortunately, vitrification has acquired the stigma that it is too expensive to receive further consideration as an alternative to solidification in high volume treatment applications. Ex situ solidification and vitrification are the competing methods for treating in excess of 450,000 m 3 of low-level radioactive and mixed waste at the Fernald Environmental Management Project (FEMP or simply, Fernald) located near Cincinnati, Ohio. This paper s a detailed study done to: compare the economics of the solidification and vitrification processes; determine if the stigma assigned to vitrification is warranted; determine if investing millions of dollars into vitrification development, along with solidification development, at Fernald is warranted. Common parameters were determined and detailed life-cycle cost estimates were made. Incorporating the unit costs into a computer spreadsheet allowed 'what if' scenarios to be performed. Some scenarios investigated included variation of: remediation times, amount of wastes treated, treatment efficiencies, electrical and material costs and escalation

  18. Mixed waste, preparing for 1996

    International Nuclear Information System (INIS)

    Duke, D.L.

    1995-01-01

    The Environmental Protection Agency has recently approved an extension to the enforcement policy for the storage of restricted mixed waste. Under this policy, EPA assigns a reduced enforcement priority to violations of the 40CFR268.50 prohibition on storage of restricted waste. Eligibility for the lower enforcement priority afforded by the policy is subject to specified conditions. The recent extension is for a two year period, and agency personnel have advised that it may be difficult to extend the enforcement policy again. This paper reviews anticipated changes in mixed waste treatment and disposal capabilities. Types of mixed waste that may be generated, or in storage, at commercial nuclear power plants are identified. This information is evaluated to determine if the two year extension in the storage enforcement policy will be adequate for the nuclear power industry to treat or dispose of the mixed waste inventories that are identified, and if not, where potential problem areas may reside. Recommendations are then made on mixed waste management strategies

  19. In situ vitrification laboratory-scale test work plan

    International Nuclear Information System (INIS)

    Nagata, P.K.; Smith, N.L.

    1991-05-01

    The Buried Waste Program was established in October 1987 to accelerate the studies needed to develop a long-term management plan for the buried mixed waste at the Radioactive Waste Management Complex at Idaho Engineering Laboratory. The In Situ Vitrification Project is being conducted in a Comprehensive Environmental Response, Compensation, and Liability Act feasibility study format to identify methods for the long-term management of mixed buried waste. To support the overall feasibility study, the situ vitrification treatability investigations are proceeding along the three parallel paths: laboratory-scale tests, intermediate field tests, and field tests. Laboratory-scale tests are being performed to provide data to mathematical modeling efforts, which, in turn, will support design of the field tests and to the health and safety risk assessment. This laboratory-scale test work plan provides overall testing program direction to meet the current goals and objectives of the in situ vitrification treatability investigation. 12 refs., 1 fig., 7 tabs

  20. Evaluation of melter technologies for vitrification of Hanford site low-level tank waste - phase 1 testing summary report

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, C.N., Westinghouse Hanford

    1996-06-27

    Following negotiation of the fourth amendment to the Tri- Party Agreement for Hanford Site cleanup, commercially available melter technologies were tested during 1994 and 1995 for vitrification of the low-level waste (LLW) stream to be derived from retrieval and pretreatment of the radioactive defense wastes stored in 177 underground tanks. Seven vendors were selected for Phase 1 testing to demonstrate vitrification of a high-sodium content liquid LLW simulant. The tested melter technologies included four Joule-heated melters, a carbon electrode melter, a combustion melter, and a plasma melter. Various dry and slurry melter feed preparation processes also were tested. The technologies and Phase 1 testing results were evaluated and a preliminary technology down-selection completed. This report describes the Phase 1 LLW melter vendor testing and the tested technologies, and summarizes the testing results and the preliminary technology recommendations.

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

    International Nuclear Information System (INIS)

    Callow, R.A.; Weidner, J.R.; Loehr, C.A.; Bates, S.O.; Thompson, L.E.; McGrail, B.P.

    1991-08-01

    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

  2. Leaching behaviour of a glassy slag and derived glass-ceramics from arc-plasma vitrification of hospital wastes

    OpenAIRE

    Romero, Maximina; Hernández, M. S.; Rincón López, Jesús María

    2009-01-01

    The arc-plasma vitrification of a hospital wastes containing metals and inorganic oxides yields to a leach-resistant glassy or vitreous slag, which can be environmentally safe for landfill disposal or could be transformed in glass-ceramic tiles with physical and mechanical properties similar to those showed by marketable products for building applications. Standard methods have been used for testing the leachability of elements from this new type of tiles. The water resistance was evaluated b...

  3. DESIGN OF THE DEMOSNTRATION BULK VITRIFICATION SYSTEM FOR THE SUPPLEMENTAL TREATMENT OF LOW ACTIVITY TANK WASTE AT HANFORD

    International Nuclear Information System (INIS)

    VAN BEEK JE

    2008-01-01

    In June 2004, the Demonstration Bulk Vitrification System (DBVS) was initiated with the intent to design, construct, and operate a full-scale bulk vitrification pilot-plant to treat low-activity tank waste from Hanford Tank 241-S-109. The DBVS facility uses In-Container Vitrification(trademark) (ICV(trademark)) at the core of the treatment process. The basic process steps combine liquid low-activity waste (LAW) and glassformers; dry the mixture; and then vitrify the mixture in a batch feed-while-melt process in a refractory lined steel container. Off-gases are processed through a state-of-the-art air pollution control system including sintered-metal filtration, thermal oxidation, acid gas scrubbing, and high-efficiency particulate air (HEPA) and high-efficiency gas adsorber (HEGA) filtration. Testing has focused on development and validation of the waste dryer, ICV, and sintered-metal filters (SMFs) equipment, operations enhancements, and glass formulation. With a parallel testing and design process, testing has allowed improvements to the DBVS equipment configuration and operating methodology, since its original inception. Design improvements include optimization of refractory panels in the ICV, simplifying glassformer addition equipment, increasing the number of waste feed chutes to the ICV, and adding capability for remote clean-out of piping, In addition, the U.S. Department of Energy (DOE) has provided an independent review of the entire DBVS process. While the review did not find any fatal flaws, some technical issues were identified that required a re-evaluation of the DBVS design and subsequent changes to the design. A 100 percent design package for the pilot plant will be completed and submitted to DOE for review in early 2008 that incorporates process improvements substantiated through testing and reviews. This paper provides a description of the bulk vitrification process and a discussion of major equipment design changes that have occurred based on full

  4. Assessing mixed waste treatment technologies

    International Nuclear Information System (INIS)

    Berry, J.B.; Bloom, G.A.; Hart, P.W.

    1994-01-01

    The US Department of Energy (DOE) is responsible for the management and treatment of its mixed low-level wastes (MLLW). As discussed earlier in this conference MLLW are regulated under both the Resource Conservation and Recovery Act and various DOE orders. During the next 5 years, DOE will manage over 1,200,000 m 3 of MLLW and mixed transuranic (MTRU) waste at 50 sites in 22 states (see Table 1). The difference between MLLW and MTRU waste is in the concentration of elements that have a higher atomic weight than uranium. Nearly all of this waste will be located at 13 sites. More than 1400 individual mixed waste streams exist with different chemical and physical matrices containing a wide range of both hazardous and radioactive contaminants. Their containment and packaging vary widely (e.g., drums, bins, boxes, and buried waste). This heterogeneity in both packaging and waste stream constituents makes characterization difficult, which results in costly sampling and analytical procedures and increased risk to workers

  5. Study on fused/cast AZS refractories for deployment in vitrification of radioactive waste effluents

    Energy Technology Data Exchange (ETDEWEB)

    Sengupta, Pranesh, E-mail: praneshsengupta@gmail.com [Materials Science Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Mishra, R.K.; Soudamini, N. [Waste Management Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Sen, D.; Mazumder, S. [Solid State Physics Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Kaushik, C.P. [Waste Management Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Ajithkumar, T.G. [Central NMR Facility, National Chemical Laboratory, Pune 411 008 (India); Banerjee, K. [Nuclear Recycle Group, Mumbai 400 085 (India)

    2015-12-15

    ‘Fused/cast Al{sub 2}O{sub 3}–ZrO{sub 2}–SiO{sub 2} (FC-AZS)’ is being considered as ‘glass contact refractory’ within ceramic melters, to be used for nuclear waste immobilization. Microstructural analyses reveal random distributions of baddeleyite (ZrO{sub 2}) within aluminosilicate (Al{sub 2}SiO{sub 5}) matrix. {sup 27}Al and {sup 29}Si NMR data suggest that within aluminosilicate matrix Al occurs in both 4- and 6-fold co-ordinations whereas Si prefers a 4-fold environment. Polydispersity of pores has been studied with small-angle neutron scattering (SANS) technique. Corrosion rates of FC-AZS within 6 M HNO{sub 3}, simulated wastes (500 h exposure), and borosilicate melt (975 °C, 800 h exposure) are found to be 0.38 × 10{sup 3} μmy{sup −1}, 0.13 × 10{sup 3} μmy{sup −1} and 4.75 × 10{sup 3} μmy{sup −1} respectively. A comparison of chemical interaction data clearly suggests that FC-AZS exhibits better chemical durability than AZC refractory (Al{sub 2}O{sub 3}–ZrO{sub 2}–Cr{sub 2}O{sub 3}, also used for similar purpose). Thermal cycling studies indicate that FC-AZS retains structural integrity (including compressive strength and density) even up to 20 cycles. - Highlights: • Vitrification of nuclear waste using AZS refractory within ceramic melter. • Microstructure of AZS refractory. • Interaction of AZS with simulated high level waste and glass.

  6. Costs of mixed low-level waste stabilization options

    International Nuclear Information System (INIS)

    Schwinkendorf, W.E.; Cooley, C.R.

    1998-01-01

    Selection of final waste forms to be used for disposal of DOE's mixed low-level waste (MLLW) depends on the waste form characteristics and total life cycle cost. In this paper the various cost factors associated with production and disposal of the final waste form are discussed and combined to develop life-cycle costs associated with several waste stabilization options. Cost factors used in this paper are based on a series of treatment system studies in which cost and mass balance analyses were performed for several mixed low-level waste treatment systems and various waste stabilization methods including vitrification, grout, phosphate bonded ceramic and polymer. Major cost elements include waste form production, final waste form volume, unit disposal cost, and system availability. Production of grout costs less than the production of a vitrified waste form if each treatment process has equal operating time (availability) each year; however, because of the lower volume of a high temperature slag, certification and handling costs and disposal costs of the final waste form are less. Both the total treatment cost and life cycle costs are higher for a system producing grout than for a system producing high temperature slag, assuming equal system availability. The treatment costs decrease with increasing availability regardless of the waste form produced. If the availability of a system producing grout is sufficiently greater than a system producing slag, then the cost of treatment for the grout system will be less than the cost for the slag system, and the life cycle cost (including disposal) may be less depending on the unit disposal cost. Treatment and disposal costs will determine the return on investment in improved system availability

  7. Design and heat transfer calculations of burial-bunker for one-stage melting converter for vitrification of high-level radioactive waste

    International Nuclear Information System (INIS)

    Pioro, L.S.; P'Yanykh, K.E.; Pioro, I.L.

    2001-01-01

    Widespread application of radioactive materials in different branches of industry, particularly in power engineering, has created a global problem in the area of ecological-disposal of radioactive waste (RAW). In general, three methods for reprocessing and disposal of RAW with high-level radionuclides are used: reservoir storage; burial in boreholes; and vitrification (solidification into glass blocks). Analysis of the recent methods of high level RAW (HLRAW) localization has shown that the most reliable method for long-term storage is vitrification. Vitrification allows to decrease by more than one order of magnitude the volume of HLRAW which is intended for long-term storage, and also to decrease leaching rates by 3-4 orders. This method includes incorporation of waste into physicochemical conglomerates during glass processing from active nuclides and neutral charging materials. Usually, this method consists of multistage processes. One-stage vitrification methods are seldom considered. (author)

  8. Mixed wasted integrated program: Logic diagram

    Energy Technology Data Exchange (ETDEWEB)

    Mayberry, J.; Stelle, S. [Science Applications International Corp., Idaho Falls, ID (United States); O`Brien, M. [Univ. of Arizona, Tucson, AZ (United States); Rudin, M. [Univ. of Nevada, Las Vegas, NV (United States); Ferguson, J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); McFee, J. [I.T. Corp., Albuquerque, NM (United States)

    1994-11-30

    The Mixed Waste Integrated Program Logic Diagram was developed to provide technical alternative for mixed wastes projects for the Office of Technology Development`s Mixed Waste Integrated Program (MWIP). Technical solutions in the areas of characterization, treatment, and disposal were matched to a select number of US Department of Energy (DOE) treatability groups represented by waste streams found in the Mixed Waste Inventory Report (MWIR).

  9. Development of the high-level waste high-temperature melter feed preparation flowsheet for vitrification process testing

    International Nuclear Information System (INIS)

    Seymour, R.G.

    1995-01-01

    High-level waste (HLW) feed preparation flowsheet development was initiated in fiscal year (FY) 1994 to evaluate alternative flowsheets for preparing melter feed for high-temperature melter (HTM) vitrification testing. Three flowsheets were proposed that might lead to increased processing capacity relative to the Hanford Waste Vitrification Plant (HWVP) and that were flexible enough to use with other HLW melter technologies. This document describes the decision path that led to the selection of flowsheets to be tested in the FY 1994 small-scale HTM tests. Feed preparation flowsheet development for the HLW HTM was based on the feed preparation flowsheet that was developed for the HWVP. This approach allowed the HLW program to build upon the extensive feed preparation flowsheet database developed under the HWVP Project. Primary adjustments to the HWVP flowsheet were to the acid adjustment and glass component additions. Developmental background regarding the individual features of the HLW feed preparation flowsheets is provided. Applicability of the HWVP flowsheet features to the new HLW vitrification mission is discussed. The proposed flowsheets were tested at the laboratory-scale at Pacific Northwest Laboratory. Based on the results of this testing and previously established criteria, a reductant-based flowsheet using glycolic acid and a nitric acid-based flowsheet were selected for the FY 1994 small-scale HTM testing

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-02-21

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  12. Simulation of radioelement volatility during the vitrification of radioactive wastes by arc plasma.

    Science.gov (United States)

    Ghiloufi, Imed

    2009-04-15

    A computer model is used to simulate the volatility of some radioelements cesium ((137)Cs), cobalt ((60)Co), and ruthenium ((106)Ru) during the radioactive wastes vitrification by thermal plasma. This model is based on the calculation of system composition using the free enthalpy minimization method, coupled with the equation of mass transfer at the reactional interface. The model enables the determination of the effects of various parameters (e.g., temperature, plasma current, and matrix composition) on the radioelement volatility. The obtained results indicate that any increase in molten bath temperature causes an increase in the cobalt volatility; while ruthenium has a less obvious behavior. It is also found that the oxygen flux in the carrier gas supports the radioelement incorporations in the containment matrix, except in the case of the ruthenium which is more volatile under an oxidizing atmosphere. For electrolyses effects, an increase in the plasma current considerably increases both the vaporization speed and the vaporized quantities of (137)Cs and (60)Co. The increase of silicon percentage in the containment matrix supports the incorporation of (60)Co and (137)Cs in the matrix. The simulation results are compared favorably to the experimental measurements obtained by emission spectroscopy.

  13. Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report

    International Nuclear Information System (INIS)

    Herborn, D.I.

    1993-11-01

    Westinghouse Hanford Company (WHC) is the Integrating Contractor for the Hanford Waste Vitrification Plant (HWVP) Project, and as such is responsible for preparation of the HWVP Preliminary Safety Analysis Report (PSAR). The HWVP PSAR was prepared pursuant to the requirements for safety analyses contained in US Department of Energy (DOE) Orders 4700.1, Project Management System (DOE 1987); 5480.5, Safety of Nuclear Facilities (DOE 1986a); 5481.lB, Safety Analysis and Review System (DOE 1986b) which was superseded by DOE order 5480-23, Nuclear Safety Analysis Reports, for nuclear facilities effective April 30, 1992 (DOE 1992); and 6430.lA, General Design Criteria (DOE 1989). The WHC procedures that, in large part, implement these DOE requirements are contained in WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual. This manual describes the overall WHC safety analysis process in terms of requirements for safety analyses, responsibilities of the various contributing organizations, and required reviews and approvals

  14. Ethanol from mixed waste paper

    International Nuclear Information System (INIS)

    Kerstetter, J.D.; Lyons, J.K.

    1991-01-01

    The technology, markets, and economics for converting mixed waste paper to ethanol in Washington were assessed. The status of enzymatic and acid hydrolysis projects were reviewed. The market for ethanol blended fuels in Washington shows room for expansion. The economics for a hypothetical plant using enzymatic hydrolysis were shown to be profitable

  15. Low-level radioactive waste, mixed low-level radioactive waste, and biomedical mixed waste

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    This document describes the proceedings of a workshop entitled: Low-Level Radioactive Waste, Mixed Low-Level Radioactive Waste, and Biomedical Mixed Waste presented by the National Low-Level Waste Management Program at the University of Florida, October 17-19, 1994. The topics covered during the workshop include technical data and practical information regarding the generation, handling, storage and disposal of low-level radioactive and mixed wastes. A description of low-level radioactive waste activities in the United States and the regional compacts is presented

  16. In-situ vitrification: a large-scale prototype for immobilizing radioactively contaminated waste

    International Nuclear Information System (INIS)

    Carter, J.G.; Buelt, J.L.

    1986-03-01

    Pacific Northwest Laboratory is developing the technology of in situ vitrification, a thermal treatment process for immobilizing radioactively contaminated soil. A permanent remedial action, the process incorporates radionuclides into a glass and crystalline form. The transportable procss consists of an electrical power system to vitrify the soil, a hood to contain gaseous effluents, an off-gas treatment system and cooling system, and a process control station. Large-scale testing of the in situ vitrification process is currently underway

  17. Temperature simulation of thermal plasma melting furnace for disposal of radioactive waste and preliminary research of vitrification formula

    International Nuclear Information System (INIS)

    Lin Peng; Lu Yonghong; Xiang Wenyuan; Chen Mingzhou; Liu Xiajie; Qin Yuxin

    2013-01-01

    Radioactive waste treatment techniques currently used in nuclear power plant increase the volume greatly and bring much pressure on final disposal; Thermal plasma treatment as a crucial technique to reduce the waste volume is introduced. How to improve the efficiency of the plasma energy is the limiting factor of concern. In this paper, the temperature field of thermal plasma melting furnace is simulated, the maximal temperature of fixed bed melting furnace is calculated (about 1445 ℃). According to the optional fire-resistant materials, the feasibility of furnace fabrication is discussed. Vitrification formulas for three typical radioactive wastes are tested with their feasibilities being analyzed then. Finally, the prospect of thermal plasma techniques of radioactive waste is discussed, and issues for future study are raised. (authors)

  18. Test Summary Report Vitrification Demonstration of an Optimized Hanford C-106/AY-102 Waste-Glass Formulation

    International Nuclear Information System (INIS)

    Goles, Ronald W.; Buchmiller, William C.; Hymas, Charles R.; MacIsaac, Brett D.

    2002-01-01

    In order to further the goal of optimizing Hanford?s HLW borosilicate flowsheet, a glass formulation effort was launched to develop an advanced high-capacity waste form exhibiting acceptable leach and crystal formation characteristics. A simulated C-106/AY-102 waste envelop inclusive of LAW pretreatment products was chosen as the subject of these nonradioactive optimization efforts. To evaluate this optimized borosilicate waste formulation under continuous dynamic vitrification conditions, a research-scale Joule-heated ceramic melter was used to demonstrate the advanced waste form?s flowsheet. The main objectives of this melter test was to evaluate (1) the processing characteristics of the newly formulated C-106/AY-102 surrogate melter-feed stream, (2) the effectiveness of sucrose as a glass-oxidation-state modifier, and (3) the impact of this reductant upon processing rates

  19. High level waste vitrification at the SRP [Savannah River Plant] (DWPF [Defense Waste Processing Facility] summary)

    International Nuclear Information System (INIS)

    Weisman, A.F.; Knight, J.R.; McIntosh, D.L.; Papouchado, L.M.

    1988-01-01

    The Savannah River Plant has been operating a nuclear fuel cycle since the early 1950's. Fuel and target elements are fabricated and irradiated to produce nuclear materials. After removal from the reactors, the fuel elements are processed to extract the products, and waste is stored. During the thirty years of operation including evaporation, about 30 million gallons of high level radioactive waste has accumulated. The Defense Waste Processing Facility (DWPF) under construction at Savannah River will process this waste into a borosilicate glass for long-term geologic disposal. The construction of the DWPF is about 70% complete; this paper will describe the status of the project, including design demonstrations, with an emphasis on the melter system. 9 figs

  20. Technology status of spray calcination--vitrification of high-level liquid waste for full-scale application

    International Nuclear Information System (INIS)

    Keeley, R.B.; Bonner, W.F.; Larson, D.E.

    1977-01-01

    Spray calcination and vitrification technology for stabilization of high-level nuclear wastes has been developed to the point that initiation of technology transfer to an industrial-sized facility could begin. This report discusses current process and equipment development status together with additional R and D studies and engineering evaluations needed. Preliminary full-scale process and equipment descriptions are presented. Technology application in a full-scale plant would blend three distinct maintenance design philosophies, depending on service life anticipated: (1) totally remote maintenance with limited viewing and handling equipment, (2) totally remote maintenance with extensive viewing and handling equipment, and (3) contact maintenance

  1. Hanford Waste Vitrification Program process development: Melt testing subtask, pilot-scale ceramic melter experiment, run summary

    International Nuclear Information System (INIS)

    Nakaoka, R.K.; Bates, S.O.; Elmore, M.R.; Goles, R.W.; Perez, J.M.; Scott, P.A.; Westsik, J.H.

    1996-03-01

    Hanford Waste Vitrification Program (HWVP) activities for FY 1985 have included engineering and pilot-scale melter experiments HWVP-11/HBCM-85-1 and HWVP-12/PSCM-22. Major objectives designated by HWVP fo these tests were to evaluate the processing characteristics of the current HWVP melter feed during actual melter operation and establish the product quality of HW-39 borosilicate glass. The current melter feed, defined during FY 85, consists of reference feed (HWVP-RF) and glass-forming chemicals added as frit

  2. Vitrification of Hanford wastes in a joule-heated ceramic melter and evaluation of resultant canisterized product

    International Nuclear Information System (INIS)

    Chapman, C.C.; Buelt, J.L.; Slate, S.C.; Katayama, Y.B.; Bunnell, L.R.

    1979-08-01

    Experience gained in the week-long vitrification test and characterization of the glass produced in the run support the following conclusions: The Hanford waste simulated in this test can be readily vitrified in a joule-heated ceramic melter. Physical properties of the molten glass were entirely compatible with melter operation. The average feed rate of 106 kg/h is high enough to make the ceramic melter a feasible piece of equipment for vitrifying Hanford wastes. The glass produced in this trial had good chemical durability, 6(10) -5 g/cm 2 -d. When one of the canisters was purposely dropped onto a steel pad, the damage was limited to deformation of the steel can in the impact area, cracking of a weld, and fracturing of glass in the immediate vicinity of the impact area. No glass was released from the canister as a result of the drop test. The results of this vitrification test support the technical feasibility of vitrifying Hanford wastes by means of a joule-heated ceramic melter. Surface area for large glass castings is equivalent to the mass median particle diameters between 4.27 cm (1.75 in.) and 8.91 cm (3.51 in.) even when allowed to cool rapidly by standing in ambient air. Large canisters (up to 0.91 m in dia) can be cast without large voids while standing in air if the fill rate is over 100 kg/h. 34 figures, 10 tables

  3. Vitrification of Three Low-Activity Radioactive Waste Streams from Hanford

    International Nuclear Information System (INIS)

    Ferrara, D.M.; Crawford, C.L.; Ha, B.C.; Bibler, N.E.

    1998-09-01

    As part of a demonstration for British Nuclear Fuels Limited, Incorporated (BNFL), the Immobilization Technology Section (ITS) of the Savannah River Technology Center (SRTC) has produced and characterized three low-activity waste (LAW) glasses from Hanford radioactive waste samples. The three LAW glasses were produced from radioactive supernate samples that had been treated by the Waste Processing Technology Section (WPTS) at SRTC to remove most of the radionuclides. These three glasses were produced by mixing the waste streams with between four and nine glass-forming chemicals in platinum/gold crucibles and heating the mixture to between 1120 and 1150 degrees C. Compositions of the resulting glass waste forms were close to the target compositions. Low concentrations of radionuclides in the LAW feed streams and, therefore, in the glass waste forms supported WPTS conclusions that pretreatment had been successful. No crystals were detected in the LAW glasses. In addition, all glass waste forms passed the leach tests that were performed. These included a 20 degrees C Product Consistency Test (PCT) and a modified version of the United States Environmental Protection Agency Toxicity Characteristic Leaching Procedure (TCLP)

  4. Addressing mixed waste in plutonium processing

    International Nuclear Information System (INIS)

    Christensen, D.C.; Sohn, C.L.; Reid, R.A.

    1991-01-01

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

  5. Vitrification of SRP waste by a slurry-fed ceramic melter

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1980-01-01

    Savannah River Plant (SRP) high-level waste (HLW) can be vitrified by feeding a slurry, instead of a calcine, to a joule-heated ceramic melter. Potential advantages of slurry feeding include (1) use of simpler equipment, (2) elimination of handling easily dispersed radioactive powder, (3) simpler process control, (4) effective mixing, (5) reduced off-gas volume, and (6) cost savings. Assessment of advantages and disadvantages of slurry feeding along with experimental studies indicate that slurry feeding is a promising way of vitrifying waste

  6. Radioactive waste combustion / vitrification under arc plasma: thermal and dynamic modelling; Combustion - vitrification de dechets radioactifs par plasma d'arc: modelisation de la thermique et de la dynamique

    Energy Technology Data Exchange (ETDEWEB)

    Barthelemy, B

    2003-07-01

    This thesis concerns the thermal and dynamic modelling for a combustion/vitrification process of surrogate radioactive waste under transferred arc plasma. The writer presents the confinement processes for radioactive waste using arc plasma and the different software used to model theses processes. This is followed by a description of our experimental equipment including a plasma arc reactor and an inductive system allowing the homogenization of glass temperature. A combustion/vitrification test is described. Thermal and material balances were discussed. The temperature fields of plasma arc and the glass frit conductivity are measured. Finally, the writer describes and clarifies the equations solved for the simulations of the electrically plasma arc and the glass melting including the thin layer of glass frit coating the crucible cold walls. The modelling results are presented in the form of spatial distribution of temperature, velocity and volume power... (author)

  7. Radioactive waste combustion-vitrification under arc plasma: thermal and dynamic modelling; Combustion - vitrification de dechets radioactifs par plasma d'arc: modelisation de la thermique et de la dynamique

    Energy Technology Data Exchange (ETDEWEB)

    Barthelemy, B

    2003-06-01

    This thesis concerns the thermal and dynamic modelling for a combustion/vitrification process of surrogate radioactive waste under transferred arc plasma. The writer presents the confinement processes for radioactive waste using arc plasma and the different software used to model theses processes. This is followed by a description of our experimental equipment including a plasma arc reactor and an inductive system allowing the homogenization of glass temperature. A combustion/vitrification test is described. Thermal and material balances were discussed. The temperature fields of plasma arc and the glass frit conductivity are measured. Finally, the writer describes and clarifies the equations solved for the simulations of the electrically plasma arc and the glass melting including the thin layer of glass frit coating the crucible cold walls. The modelling results are presented in the form of spatial distribution of temperature, velocity and voluminal power... (author)

  8. Preliminary hazards analysis -- vitrification process

    International Nuclear Information System (INIS)

    Coordes, D.; Ruggieri, M.; Russell, J.; TenBrook, W.; Yimbo, P.

    1994-06-01

    This paper presents a Preliminary Hazards Analysis (PHA) for mixed waste vitrification by joule heating. The purpose of performing a PHA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PHA is then followed by a Preliminary Safety Analysis Report (PSAR) performed during Title 1 and 2 design. The PSAR then leads to performance of the Final Safety Analysis Report performed during the facility's construction and testing. It should be completed before routine operation of the facility commences. This PHA addresses the first four chapters of the safety analysis process, in accordance with the requirements of DOE Safety Guidelines in SG 830.110. The hazards associated with vitrification processes are evaluated using standard safety analysis methods which include: identification of credible potential hazardous energy sources; identification of preventative features of the facility or system; identification of mitigative features; and analyses of credible hazards. Maximal facility inventories of radioactive and hazardous materials are postulated to evaluate worst case accident consequences. These inventories were based on DOE-STD-1027-92 guidance and the surrogate waste streams defined by Mayberry, et al. Radiological assessments indicate that a facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous materials assessment indicates that a Mixed Waste Vitrification facility will be a Low Hazard facility having minimal impacts to offsite personnel and the environment

  9. Preliminary hazards analysis -- vitrification process

    Energy Technology Data Exchange (ETDEWEB)

    Coordes, D.; Ruggieri, M.; Russell, J.; TenBrook, W.; Yimbo, P. [Science Applications International Corp., Pleasanton, CA (United States)

    1994-06-01

    This paper presents a Preliminary Hazards Analysis (PHA) for mixed waste vitrification by joule heating. The purpose of performing a PHA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PHA is then followed by a Preliminary Safety Analysis Report (PSAR) performed during Title 1 and 2 design. The PSAR then leads to performance of the Final Safety Analysis Report performed during the facility`s construction and testing. It should be completed before routine operation of the facility commences. This PHA addresses the first four chapters of the safety analysis process, in accordance with the requirements of DOE Safety Guidelines in SG 830.110. The hazards associated with vitrification processes are evaluated using standard safety analysis methods which include: identification of credible potential hazardous energy sources; identification of preventative features of the facility or system; identification of mitigative features; and analyses of credible hazards. Maximal facility inventories of radioactive and hazardous materials are postulated to evaluate worst case accident consequences. These inventories were based on DOE-STD-1027-92 guidance and the surrogate waste streams defined by Mayberry, et al. Radiological assessments indicate that a facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous materials assessment indicates that a Mixed Waste Vitrification facility will be a Low Hazard facility having minimal impacts to offsite personnel and the environment.

  10. Defense Waste Processing Facility (DWPF): The vitrification of high-level nuclear waste. (Latest citations from the Bibliographic database). Published Search

    International Nuclear Information System (INIS)

    1993-09-01

    The bibliography contains citations concerning a production-scale facility and the world's largest plant for the vitrification of high-level radioactive nuclear wastes (HLW) located in the United States. Initially based on the selection of borosilicate glass as the reference waste form, the citations present the history of the development including R ampersand D projects and the actual construction of the production facility at the DOE Savannah River Plant (SRP). (Contains a minimum of 177 citations and includes a subject term index and title list.)

  11. Modeling in situ vitrification

    International Nuclear Information System (INIS)

    Mecham, D.C.; MacKinnon, R.J.; Murray, P.E.; Johnson, R.W.

    1990-01-01

    In Situ Vitrification (ISV) process is being assessed by the Idaho National Engineering Laboratory (INEL) to determine its applicability to transuranic and mixed wastes buried at INEL'S Subsurface Disposal Area (SDA). This process uses electrical resistance heating to melt waste and contaminated soil in place to produce a durable glasslike material that encapsulates and immobilizes buried wastes. This paper outlines the requirements for the model being developed at the INEL which will provide analytical support for the ISV technology assessment program. The model includes representations of the electric potential field, thermal transport with melting, gas and particulate release, vapor migration, off-gas combustion and process chemistry. The modeling objectives are to help determine the safety of the process by assessing the air and surrounding soil radionuclides and chemical pollution hazards, the nuclear criticality hazard, and the explosion and fire hazards, help determine the suitability of the ISV process for stabilizing the buried wastes involved, and help design laboratory and field tests and interpret results. 3 refs., 2 figs., 1 tab

  12. Application of DOE prescribed guides to the evaluation of Hanford's Mixed Low Level Solid Waste Treatment Options

    International Nuclear Information System (INIS)

    Campbell, B.F.; Nash, C.R.

    1994-10-01

    A recent Westinghouse Hanford Company report (WHC-SD-W100-ES-008, February, 1994), compared a Vitrification process to the WRAP-2A Grout/PE process for the treatment of Mixed Low Level Waste (MLLW). This comparison applied a limited scope numerical evaluation to compare technology complexity of the two processes, but focused primarily on capital and operating costs. The work reported here is supplementary to WHC-SD-Wl00-ES-008. It provides a record of the application of the more formal DOE-prescribed criteria (Treatment Selection Guides for Federal Facility Compliance Act Draft Site Treatment Plans) to the Vitrification and Grout/PE processes previously evaluated. Results of the evaluation favored the Grout/PE process by a weighted score of 83 to 78 over the Plasma arc vitrification process

  13. The role of Tetronics plasma vitrification technology in the management and conditioning of nuclear waste

    International Nuclear Information System (INIS)

    Deegan, David; Scales, Charlie

    2007-01-01

    Plasma Arc Technology is finding wider application in the treatment of hazardous waste materials an area which has a lot of synergy with radioactive waste management. It is being stimulated by the increasing demands of regulatory and economic drivers; currently, within the Integrated Waste Management (IWM) sector, there is a climate of rising costs, limited numbers of technological solutions, restricted access to traditional disposal based solutions and a significant levels of market consolidation. Traditionally, the IWM sector has operated with basic mixing technology solutions: e.g. physiochemical consolidation, physiochemical separation, neutralisation and basic material bulking, with ultimate reliance on landfill, cement based encapsulation and high temperature incineration (HTI). The impact of national statutes, the value of national liabilities and infra-structural deficiencies is demanding constant technological advancement for continued regulatory compliance. This paper presents information on Tetronics' plasma based solution, for the treatment of Asbestos Containing Materials (ACM) and Plutonium Containing Material (PCM). (authors)

  14. Qualification and characterization programmes for disposal of a glass product resulting from high level waste vitrification in the PAMELA installation of BELGOPROCESS

    International Nuclear Information System (INIS)

    Goeyse, A. de; De, A.K.; Demonie, M.; Iseghem, P. van

    1993-01-01

    In the framework of a general quality assurance and quality control (QA/QC) programme, the quality of a conditioned waste product is achieved in two phases. The first phase is the design of a process and facility which will ensure the required quality of the product. In the second phase the conformance of the product with the preset requirements is verified. NIRAS/ONDRAF, as the agency responsible for the management of all radioactive waste in Belgium (including treatment, conditioning, storage and disposal), controls compliance with the quality requirements during both phases. The purpose of the paper is to describe the different phases of this general procedure in the case of a vitrified HLW product resulting from a vitrification campaign in the PAMELA facility at the BELGOPROCESS site. The active glass product of type SM527 produced during the vitrification of highly enriched waste concentrate (HEWC) (resulting from the reprocessing of highly enriched uranium fuel) has been selected for illustration. During the process qualification phase, the Deutsche Gesellschaft fuer Wiederaufarbeitung von Kernbrennstoffen mbH, responsible for the development of the vitrification process of PAMELA, defined and performed and R and D programmed for each glass product originating from the vitrification of the different HEWC solutions stored at the BELGOPROCESS site. At the end of this qualification phase a data catalogue was prepared. In order to ensure that the active glass product corresponds with the selected product from the data catalogue, the QA/QC handbook for the vitrification process describes all measures to be taken by the waste producer, BELGOPROCESS, during the vitrification. Finally, verification analyses are performed by the characterization of inactive and active samples by an independent laboratory. This phase is called the product quality verification phase. The details of the characterization programmes performed during the different phases and their results

  15. Tank waste remediation system high-level waste vitrification system development and testing requirements

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-01-01

    This document provides the fiscal year (FY) 1995 recommended high-level waste melter system development and testing (D and T) requirements. The first phase of melter system testing (FY 1995) will focus on the feasibility of high-temperature operation of recommended high-level waste melter systems. These test requirements will be used to establish the basis for defining detailed testing work scope, cost, and schedules. This document includes a brief summary of the recommended technologies and technical issues associated with each technology. In addition, this document presents the key D and T activities and engineering evaluations to be performed for a particular technology or general melter system support feature. The strategy for testing in Phase 1 (FY 1995) is to pursue testing of the recommended high-temperature technologies, namely the high-temperature, ceramic-lined, joule-heated melter, referred to as the HTCM, and the high-frequency, cold-wall, induction-heated melter, referred to as the cold-crucible melter (CCM). This document provides a detailed description of the FY 1995 D and T needs and requirements relative to each of the high-temperature technologies

  16. Hanford Site annual dangerous waste report: Volume 4, Waste Management Facility report, Radioactive mixed waste

    International Nuclear Information System (INIS)

    1994-01-01

    This report contains information on radioactive mixed wastes at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, handling method and containment vessel, waste number, waste designation and amount of waste

  17. Hanford Site annual dangerous waste report: Volume 2, Generator dangerous waste report, radioactive mixed waste

    International Nuclear Information System (INIS)

    1994-01-01

    This report contains information on radioactive mixed wastes at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, waste number, waste designation, weight, and waste designation

  18. Mercury removal from solid mixed waste

    International Nuclear Information System (INIS)

    Gates, D.D.; Morrissey, M.; Chava, K.K.; Chao, K.

    1994-01-01

    The removal of mercury from mixed wastes is an essential step in eliminating the temporary storage of large inventories of mixed waste throughout the Department of Energy (DOE) complex. Currently thermal treatment has been identified as a baseline technology and is being developed as part of the DOE Mixed Waste Integrated Program (MWIP). Since thermal treatment will not be applicable to all mercury containing mixed waste and the removal of mercury prior to thermal treatment may be desirable, laboratory studies have been initiated at Oak Ridge National Laboratory (ORNL) to develop alternative remediation technologies capable of removing mercury from certain mixed waste. This paper describes laboratory investigations of the KI/I 2 leaching processes to determine the applicability of this process to mercury containing solid mixed waste

  19. Commercial mixed waste treatment and disposal

    International Nuclear Information System (INIS)

    Vance, J.K.

    1994-01-01

    At the South Clive, Utah, site, Envirocare of Utah, Inc., (Envirocare), currently operates a commercial low-activity, low-level radioactive waste facility, a mixed waste RCRA Part B storage and disposal facility, and an 11e.(2) disposal facility. Envirocare is also in the process of constructing a Mixed Waste Treatment Facility. As the nation's first and only commercial treatment and disposal facility for such waste, the information presented in this segment will provide insight into their current and prospective operations

  20. Design and operating features of the high-level waste vitrification system for the West Valley demonstration project

    International Nuclear Information System (INIS)

    Siemens, D.H.; Beary, M.M.; Barnes, S.M.; Berger, D.N.; Brouns, R.A.; Chapman, C.C.; Jones, R.M.; Peters, R.D.; Peterson, M.E.

    1986-03-01

    A liquid-fed joule-heated ceramic melter system is the reference process for immobilization of the high-level liquid waste in the US and several foreign countries. This system has been under development for over ten years at Pacific Northwest Laboratory and other national laboratories operated for the US Department of Energy. Pacific Northwest Laboratory contributed to this research through its Nuclear Waste Treatment Program and used applicable data to design and test melters and related systems using remote handling of simulated radioactive wastes. This report describes the equipment designed in support of the high-level waste vitrification program at West Valley, New York. Pacific Northwest Laboratory worked closely with West Valley Nuclear Services Company to design a liquid-fed ceramic melter, a liquid waste preparation and feed tank and pump, an off-gas treatment scrubber, and an enclosed turntable for positioning the waste canisters. Details of these designs are presented including the rationale for the design features and the alternatives considered

  1. Robotics for mixed waste operations, demonstration description

    International Nuclear Information System (INIS)

    Ward, C.R.

    1993-01-01

    The Department of Energy (DOE) Office of Technology Development (OTD) is developing technology to aid in the cleanup of DOE sites. Included in the OTD program are the Robotics Technology Development Program and the Mixed Waste Integrated Program. These two programs are working together to provide technology for the cleanup of mixed waste, which is waste that has both radioactive and hazardous constituents. There are over 240,000 cubic meters of mixed low level waste accumulated at DOE sites and the cleanup is expected to generate about 900,000 cubic meters of mixed low level waste over the next five years. This waste must be monitored during storage and then treated and disposed of in a cost effective manner acceptable to regulators and the states involved. The Robotics Technology Development Program is developing robotics technology to make these tasks safer, better, faster and cheaper through the Mixed Waste Operations team. This technology will also apply to treatment of transuranic waste. The demonstration at the Savannah River Site on November 2-4, 1993, showed the progress of this technology by DOE, universities and industry over the previous year. Robotics technology for the handling, characterization and treatment of mixed waste as well robotics technology for monitoring of stored waste was demonstrated. It was shown that robotics technology can make future waste storage and waste treatment facilities better, faster, safer and cheaper

  2. Formulation of special glass frit and its use for decontamination of Joule melter employed for vitrification of high level and radioactive liquid waste

    International Nuclear Information System (INIS)

    Valsala, T.P.; Mishra, P.K.; Thakur, D.A.; Ghongane, D.E.; Jayan, R.V.; Dani, U.; Sonavane, M.S.; Kulkarni, Y.

    2012-01-01

    Advanced vitrification system at TWMP Tarapur was used for successful vitrification of large volume of HLW stored in waste tank farm. After completion of the operational life of the joule melter, dismantling was planned. Prior to the dismantling, the hold up inventory of active glass product from the melter was flushed out using specially formulated inactive glass frit to reduce the air activity buildup in the cell during dismantling operations. The properties of the special glass frit prepared are comparable with that of the regular product glass. More than 94% of holdup activity was flushed out from the joule melter prior to the dismantling of the melter. (author)

  3. DC Graphite Arc Melter for vitrification of low-level waste

    International Nuclear Information System (INIS)

    Desrosiers, A.E.; Wilver, P.J.; Wittle, J.K.

    1996-01-01

    The volume of mixed waste continues to increase with few options for its permanent disposal other than storage on site. This mixed waste is being generated by not only the Department of Energy at government sites but by the private sector in hospitals and at electrical utility sites. Bartlett Services, Inc. proposes to offer a service to treat these materials to both reduce the volume and stabilize the radionuclides in a vitrified material. This product will be formed in the DC Graphite Arc Melters developed by Electro-Pyrolysis, Inc. and being offered for commercial design, sale and installation by Svedala Industries, Pyro Division. The process is a high temperature procedure which pyrolytically decomposes the organic portion of the waste to form clean hydrogen and carbon monoxide and solid carbon. The inorganic portion, containing the radioactive components, melts to produce a stable glass which is resistant to environmental leaching and will remain stable until the radioactivity has decreased to a safe level. Glasses produced with surrogate materials such as cesium and cerium have been shown to pass the Product Compatibility Test (PCT). The process being proposed for this treatment utilizes a sealed melter system having the capability of melting wastes containing both metallic and inorganic materials. This process, unlike joule heated melters, is capable of operating to temperatures of 1600 degrees C or higher. Since the system is heated electrically, oxidation is not required to create the heat. Since the system is pyrolytic, relatively small quantities of gas are produced. These gases may have beneficial uses in producing chemicals or may be used as a clean fuel

  4. Study of the ideal vitrification conditions of sodium containing waste after oxidation on a fluidized bed

    International Nuclear Information System (INIS)

    Petitfour, B.; Rahier, A.

    1997-08-01

    The Belgian Nuclear Research Centre SCK-CEN has patented a new process for the treatment of metallic sodium. This process is fully integrated since it allows to vitrify easily the mixture resulting from the oxidation step. To ensure the link between the treatment and the conditioning, the vitrification conditions have been studied. It is confirmed that an adequate control of the temperature decrease during the vitrification is essential to obtain a product whose area is well known and controlled. Also, the release of Cs and Co has been examined through leaching tests. The study has led to the choice of adequate composition ranges for SiO 2 , Na 2 O, Al 2 O 3 , B 2 O 3 and CaO. Further studies will be carried out to assess the possible use of vitro-crystalline materials for long term conditioning. vitro-crystalline materials for long term conditioning

  5. Denitration and chemical precipitation of medium level liquid wastes and conditioning of high level wastes from low level liquid wastes by a roll dryer and subsequent vitrification

    International Nuclear Information System (INIS)

    Halaszovich, S.; Dix, S.; Harms, R.

    1987-01-01

    Medium level liquid waste (MAW) from the reprocessing need after being fixed in cement an additional shielding to meet required radiation limits for handling and transportation. Normally this shielding consists of concrete and its weight and volume is several times higher than that of the waste product itself. By means of caesium separation using nickel-potassium-hexacyanoferrate and after few years of interim storage waiting for the decay of Ruthenium and Antimony the activities will be reduced below permissible values. (13 MBq/l in waste solution for Cs, 28 MBq/l for Sb and 34 MBq/l for Ru). Below these limits there is no need for additional shielding after cementation in a 400 l drum. Experimental results show, that Caesium can be precipitated and separated effectively not only in laboratory but also in a larger scale under hot cell conditions. The process investigated in this work has been developed from the FIPS process for vitrification of highly radioactive fission product solutions. It consists of: denitration, precipitation, sludge separation, drying and melting

  6. The mixed waste management facility

    International Nuclear Information System (INIS)

    Streit, R.D.

    1995-10-01

    During FY96, the Mixed Waste Management Facility (MWMF) Project has the following major objectives: (1) Complete Project Preliminary Design Review (PDR). (2) Complete final design (Title II) of MWMF major systems. (3) Coordinate all final interfaces with the Decontamination and Waste Treatment Facility (DWTF) for facility utilities and facility integration. (4) Begin long-lead procurements. (5) Issue Project Baseline Revision 2-Preliminary Design (PB2), modifying previous baselines per DOE-requested budget profiles and cost reduction. Delete Mediated Electrochemical Oxidation (MEO) as a treatment process for initial demonstration. (6) Complete submittal of, and ongoing support for, applications for air permit. (7) Begin detailed planning for start-up, activation, and operational interfaces with the Laboratory's Hazardous Waste Management Division (HWM). In achieving these objectives during FY96, the Project will incorporate and implement recent DOE directives to maximize the cost savings associated with the DWTF/MWMF integration (initiated in PB1.2); to reduce FY96 new Budget Authority to ∼$10M (reduced from FY97 Validation of $15.3M); and to keep Project fiscal year funding requirements largely uniform at ∼$10M/yr. A revised Project Baseline (i.e., PB2), to be issued during the second quarter of FY96, will address the implementation and impact of this guidance from an overall Project viewpoint. For FY96, the impact of this guidance is that completion of final design has been delayed relative to previous baselines (resulting from the delay in the completion of preliminary design); ramp-up in staffing has been essentially eliminated; and procurements have been balanced through the Project to help balance budget needs to funding availability

  7. Innovative technology summary report: Transportable vitrification system

    International Nuclear Information System (INIS)

    1998-09-01

    At the end of the cold war, many of the Department of Energy's (DOE's) major nuclear weapons facilities refocused their efforts on finding technically sound, economic, regulatory compliant, and stakeholder acceptable treatment solutions for the legacy of mixed wastes they had produced. In particular, an advanced stabilization process that could effectively treat the large volumes of settling pond and treatment sludges was needed. Based on this need, DOE and its contractors initiated in 1993 the EM-50 sponsored development effort required to produce a deployable mixed waste vitrification system. As a consequence, the Transportable Vitrification System (TVS) effort was undertaken with the primary requirement to develop and demonstrate the technology and associated facility to effectively vitrify, for compliant disposal, the applicable mixed waste sludges and solids across the various DOE complex sites. After 4 years of development testing with both crucible and pilot-scale melters, the TVS facility was constructed by Envitco, evaluated and demonstrated with surrogates, and then successfully transported to the ORNL ETTP site and demonstrated with actual mixed wastes in the fall of 1997. This paper describes the technology, its performance, the technology applicability and alternatives, cost, regulatory and policy issues, and lessons learned

  8. Mixed Waste Focus Area program management plan

    International Nuclear Information System (INIS)

    Beitel, G.A.

    1996-10-01

    This plan describes the program management principles and functions to be implemented in the Mixed Waste Focus Area (MWFA). The mission of the MWFA is to provide acceptable technologies that enable implementation of mixed waste treatment systems developed in partnership with end-users, stakeholders, tribal governments and regulators. The MWFA will develop, demonstrate and deliver implementable technologies for treatment of mixed waste within the DOE Complex. Treatment refers to all post waste-generation activities including sampling and analysis, characterization, storage, processing, packaging, transportation and disposal

  9. Investigation of potential for occurrence of molten soil displacement events during in situ vitrification of combustible wastes

    International Nuclear Information System (INIS)

    Roberts, J.S.; Strachan, C.W.; Luey, J.

    1993-02-01

    Computer simulations are used to investigate the application of in situ vitrification (ISV) for processing contaminated soil containing high loadings of solid, compressible waste material, typical of landfills and solid waste trenches. Specifically, these simulations predict whether significant displacement of molten soil, due to large, 1 to 2 m diameter, gas bubbles rising up through the ISV melt, are likely to occur during processing of combustible waste-loaded sites. These bubbles are believed to originate from high-pressure regions below the melt caused by vaporization of water and gases generated by the combustion, volatilization, or pyrolyzation of combustible materials in the waste. Simulations were run using the TOUGH2 computer code to predict pressures underneath the ISV melt TOUGH2 is an unsaturated groundwater modeling code capable of treating non-isothermal problems. These simulations include moving melt front and simple pyrolysis models and investigate how the gas pressure in the soil below the melt is affected by melt progression rate, soil permeability, combustible and impermeable material loading. The following, conclusions have been drawn based on the TOUGH2 simulations

  10. Mixed Waste Integrated Program: A technology assessment for mercury-containing mixed wastes

    International Nuclear Information System (INIS)

    Perona, J.J.; Brown, C.H.

    1993-03-01

    The treatment of mixed wastes must meet US Environmental Protection Agency (EPA) standards for chemically hazardous species and also must provide adequate control of the radioactive species. The US Department of Energy (DOE) Office of Technology Development established the Mixed Waste Integrated Program (MWIP) to develop mixed-waste treatment technology in support of the Mixed Low-Level Waste Program. Many DOE mixed-waste streams contain mercury. This report is an assessment of current state-of-the-art technologies for mercury separations from solids, liquids, and gases. A total of 19 technologies were assessed. This project is funded through the Chemical-Physical Technology Support Group of the MWIP

  11. Experiences with treatment of mixed waste

    Energy Technology Data Exchange (ETDEWEB)

    Dziewinski, J.; Marczak, S.; Smith, W.H. [Los Alamos National Lab., NM (United States); Nuttall, E. [Univ. of New Mexico, Albuquerque, NM (United States). Chemical and Nuclear Engineering Dept.

    1996-04-10

    During its many years of research activities involving toxic chemicals and radioactive materials, Los Alamos National Laboratory (Los Alamos) has generated considerable amounts of waste. Much of this waste includes chemically hazardous components and radioisotopes. Los Alamos chose to use an electrochemical process for the treatment of many mixed waste components. The electro-chemical process, which the authors are developing, can treat a great variety of waste using one type of equipment built at a moderate expense. Such a process can extract heavy metals, destroy cyanides, dissolve contamination from surfaces, oxidize toxic organic compounds, separate salts into acids and bases, and reduce the nitrates. All this can be accomplished using the equipment and one crew of trained operating personnel. Results of a treatability study of chosen mixed wastes from Los Alamos Mixed Waste Inventory are presented. Using electrochemical methods cyanide and heavy metals bearing wastes were treated to below disposal limits.

  12. Experiences with treatment of mixed waste

    International Nuclear Information System (INIS)

    Dziewinski, J.; Marczak, S.; Smith, W.H.; Nuttall, E.

    1996-01-01

    During its many years of research activities involving toxic chemicals and radioactive materials, Los Alamos National Laboratory (Los Alamos) has generated considerable amounts of waste. Much of this waste includes chemically hazardous components and radioisotopes. Los Alamos chose to use an electrochemical process for the treatment of many mixed waste components. The electro-chemical process, which the authors are developing, can treat a great variety of waste using one type of equipment built at a moderate expense. Such a process can extract heavy metals, destroy cyanides, dissolve contamination from surfaces, oxidize toxic organic compounds, separate salts into acids and bases, and reduce the nitrates. All this can be accomplished using the equipment and one crew of trained operating personnel. Results of a treatability study of chosen mixed wastes from Los Alamos Mixed Waste Inventory are presented. Using electrochemical methods cyanide and heavy metals bearing wastes were treated to below disposal limits

  13. The KS-KT-100 plant for two-stage vitrification of radioactive waste: results of tests with simulators

    International Nuclear Information System (INIS)

    Davydov, V.I.; Dobrygin, P.G.; Dolgov, V.V.; Sergeev, G.A.

    1976-01-01

    The Soviet Union has developed a two-stage process for phosphate vitrification of liquid radioactive waste involving the use, at the initial stage, of calcination in the pseudo-liquefied layer, followed by melting of the calcinate in a ceramic crucible (second stage). On the basis of the laboratory studies and bench tests using experimental equipment, the authors have developed and tried out an enlarged plant - the KS-KT-100. The plant includes units for preparing the solution, evaporation, calcination, melting and gas purification. The initial solution containing 240 g/litre of aluminium nitrate, 125 g/litre of sodium nitrate, 120 to 130 g/litre of orthophosphoric acid, and 90 to 150 g/litre of industrial molasses simulated fluxed nitrate waste. The tests have shown that the various units operate satisfactorily. The authors have determined the technological parameters for evaporation, calcination of the solution and melting of the calcinate. The presence of molasses in the solution (150 g/litre) makes it possible to decompose and distil 40% of the nitrate ion during evaporation. The calcination temperature is 350 to 400 0 C, and the fluidization rate 1.5 m/s. The capacity of the plant for the initial solution is 100 litres/h, for the evaporated solution 65 litres/h, and for the glass 20 kg/h. The efficiency of the gas purification system ranges between 10 7 and 10 9 . The test results show the feasibility of the two-stage method of vitrification in actual practice. (author)

  14. Challenges in development of matrices for vitrification of old legacy waste and high-level radioactive waste generated from reprocessing of AHWR and FBR spent fuel

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2012-01-01

    Majority of radioactivity in entire nuclear fuel cycle is concentrated in HLW. A three step strategy for management of HLW has been adopted in India. This involves immobilization of waste oxides in stable and inert solid matrices, interim retrievable storage of the conditioned waste product under continuous cooling and disposal in deep geological formations. Glass has been accepted as most suitable matrix world-wide for immobilization of HLW, because of its attractive features like ability to accommodate wide range of waste constituents, modest processing temperatures, adequate chemical, thermal and radiation stability. Borosilicate glass matrix developed by BARC in collaboration with CGCRI has been adopted in India for immobilization of HLW. In view of compositional variation of HLW from site to site, tailor make changes in the glass formulations are often necessary to incorporate all the waste constituents and having the product of desirable characteristics. The vitrified waste products made with different glass formulations and simulated waste need to be characterized for chemical durability, thermal stability, homogeneity etc. before finalizing a suitable glass formulation. The present extended abstract summarises the studies carried out for development of glass formulations for vitrification of legacy waste and futuristic waste likely to be generated from AHWR and FBR having wide variations in their compositions. The presently stored HLW at Trombay is characterized by significant concentrations of uranium, sodium and sulphate in addition to fission products, corrosion products and small amount of other actinides

  15. Test plan for evaluation of plasma melter technology for vitrification of high-sodium content low-level radioactive liquid wastes

    International Nuclear Information System (INIS)

    McLaughlin, D.F.; Lahoda, E.J.; Gass, W.R.; D'Amico, N.

    1994-01-01

    This document provides a test plan for the conduct of plasma arc vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. The vendor providing this test plan and conducting the work detailed within it [one of seven selected for glass melter testing under Purchase Order MMI-SVV-384212] is the Westinghouse Science and Technology Center (WSTC) in Pittsburgh, PA. WSTC authors of the test plan are D. F. McLaughlin, E. J. Lahoda, W. R. Gass, and N. D'Amico. The WSTC Program Manager for this test is D. F. McLaughlin. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes melting of glass frit with Hanford LLW Double-Shell Slurry Feed waste simulant in a plasma arc fired furnace

  16. In-situ vitrification: pilot-scale development

    International Nuclear Information System (INIS)

    Timmerman, C.L.; Brouns, R.A.; Buelt, J.L.; Oma, K.H.

    1983-01-01

    Pacific Northwest Laboratory (PNL) is developing in-situ vitrification (ISV) as an in-place stabilization technique for buried radioactive and hazardous chemical wastes. The process melts the wastes and surrounding soil to produce a durable glass and crystalline waste form. These in situ vitrification process development testing and product evaluation studies are being conducted for the U.S. Department of Energy. This report discusses the results of four ISV pilot-scale field tests simulating radioactive and hazardous waste site conditions. The primary objectives of the field tests were to: demonstrate process scale-up from engineering-scale laboratory tests; verify equipment performance of the power system, electrodes and off-gas system; characterize the behavior of simulated wastes in the vitrified soil; identify waste losses to the off-gas system; and evaluate waste form durability. Test results have been encouraging. Process scaleup has been successfully demonstrated, with equipment and electrode performance equally as successful. The off-gas system effectively contained any volatile or entrained hazardous species. Vitrified soil analysis also indicated effective containment and a homogeneous distribution of nonradioactive radionuclide and hazardous waste simulants due to convective mixing during vitrification. Waste form leaching studies revealed that the ISV product has a durability similar to Pyrex glass

  17. Vitrification of nanotoxic waste (Ru) from the production of nano-catalysts for direct ethanol fuel cells

    International Nuclear Information System (INIS)

    Silva, A.C.; Julio-Junior, O.; Mello-Castanho, S.R.H.

    2010-01-01

    Nanostructured catalysts have been developed for ethanol directly use in fuel cells, which due to the economic advantages that should have widespread use in the near future. The catalysts for these devices using nano-structured metal are based, where the toxic nature and environmental risks presented by these metals are largely enhanced by nano-dispersion. Thus, the production of nano-catalysts are potentially generating highly hazardous waste for public health and the environment. This study presents the treatment and inertization of ruthenium (Ru) nanoparticles waste containing by the vitrification technique and consequent attainment of silicate glasses for potential commercial use. Compositions were prepared containing up to about 20 wt % of nano-waste by changing the basic composition of glass soda-lime-borosilicate. After the fusion, at a temperature of 1100 deg C, the glasses were characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM) and Infra-red in the Fourier transform (FT-IR) techniques. The chemical stability was evaluated by hydrolytic attack test. The glass containing 20 wt % of nano-residue showed a high chemical stability, similar to a usual soda-lime glass. (author)

  18. A summary report on feed preparation offgas and glass redox data for Hanford waste vitrification plant: Letter report

    International Nuclear Information System (INIS)

    Merz, M.D.

    1996-03-01

    Tests to evaluate feed processing options for the Hanford Waste Vitrification Plant (HWVP) were conducted by a number of investigators, and considerable data were acquired for tests of different scale, including recent full-scale tests. In this report, a comparison was made of the characteristics of feed preparation observed in tests of scale ranging from 57 ml to full-scale of 28,000 liters. These tests included Pacific Northwest Laboratory (PNL) laboratory-scale tests, Kernforschungszentrums Karlsruhe (KfK) melter feed preparation, Research Scale Melter (RSM) feed preparation, Integrated DWPF Melter System (IDMS) feed preparation, Slurry Integrated Performance Testing (SIPT) feed preparation, and formic acid addition to Hanford Neutralized Current Acid Waste (NCAW) care samples.' The data presented herein were drawn mainly from draft reports and include system characteristics such as slurry volume and depth, sweep gas flow rate, headspace, and heating and stirring characteristics. Operating conditions such as acid feed rate, temperature, starting pH, final pH, quantities and type of frit, nitrite, nitrate, and carbonate concentrations, noble metal content, and waste oxide loading were tabulated. Offgas data for CO 2 , NO x , N 2 O, NO 2 , H 2 and NH 3 were tabulated on a common basis. Observation and non-observation of other species were also noted

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-07

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

  20. Electrochemical treatment of mixed and hazardous waste

    International Nuclear Information System (INIS)

    Dziewinski, J.; Marczak, S.; Smith, W.; Nuttall, E.

    1995-01-01

    Los Alamos National Laboratory (LANL) and The University of New Mexico are jointly developing an electrochemical process for treating hazardous and radioactive wastes. The wastes treatable by the process include toxic metal solutions, cyanide solutions, and various organic wastes that may contain chlorinated organic compounds. The main component of the process is a stack of electrolytic cells with peripheral equipment such as a rectifier, feed system, tanks with feed and treated solutions, and a gas-venting system. During the treatment, toxic metals are deposited on the cathode, cyanides are oxidized on the anode, and organic compounds are anodically oxidized by direct or mediated electrooxidation, depending on their type. Bench scale experimental studies have confirmed the feasibility of applying electrochemical systems to processing of a great variety of hazardous and mixed wastes. The operating parameters have been defined for different waste compositions using surrogate wastes. Mixed wastes are currently treated at bench scale as part of the treatability study

  1. Hanford's Radioactive Mixed Waste Disposal Facility

    International Nuclear Information System (INIS)

    McKenney, D.E.

    1995-01-01

    The Radioactive Mixed Waste Disposal Facility, is located in the Hanford Site Low-Level Burial Grounds and is designated as Trench 31 in the 218-W-5 Burial Ground. Trench 31 is a Resource Conservation and Recovery Act compliant landfill and will receive wastes generated from both remediation and waste management activities. On December 30, 1994, Westinghouse Hanford Company declared readiness to operate Trench 31, which is the Hanford Site's (and the Department of Energy complex's) first facility for disposal of low-level radioactive mixed wastes

  2. Treatment of M-area mixed wastes at the Savannah River Site

    International Nuclear Information System (INIS)

    1994-06-01

    The Department of Energy has prepared this environmental assessment, DOE/EA-0918, to assess the potential environmental impacts of the treatment of mixed wastes currently stored in the M-Area at the Savannah River Site, near Aiken, South Carolina. DOE is proposing to treat and stabilize approximately 700,000 gallons of mixed waste currently stored in the Interim Treatment/Storage Facility (IT/SF) and Mixed Waste Storage Shed (MWSS). This waste material is proposed to be stabilized using a vitrification process and temporarily stored until final disposal is available by the year 2005. This document has been prepared to assess the potential environmental impacts attributable to the treatment and stabilization of M-area mixed wastes, the closure of the interim storage area, and storage of the vitrified waste until disposal in onsite RCRA vaults. Based on the analyses in the environmental assessment, the Department of Energy has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, the preparation of an environmental impact statement is not required, and the Department of Energy is issuing this finding of no significant impact

  3. Comparative life-cycle cost analysis for low-level mixed waste remediation alternatives

    International Nuclear Information System (INIS)

    Jackson, J.A.; White, T.P.; Kloeber, J.M.; Toland, R.J.; Cain, J.P.; Buitrago, D.Y.

    1995-03-01

    The purpose of this study is two-fold: (1) to develop a generic, life-cycle cost model for evaluating low-level, mixed waste remediation alternatives, and (2) to apply the model specifically, to estimate remediation costs for a site similar to the Fernald Environmental Management Project near Cincinnati, OH. Life-cycle costs for vitrification, cementation, and dry removal process technologies are estimated. Since vitrification is in a conceptual phase, computer simulation is used to help characterize the support infrastructure of a large scale vitrification plant. Cost estimating relationships obtained from the simulation data, previous cost estimates, available process data, engineering judgment, and expert opinion all provide input to an Excel based spreadsheet for generating cash flow streams. Crystal Ball, an Excel add-on, was used for discounting cash flows for net present value analysis. The resulting LCC data was then analyzed using multi-attribute decision analysis techniques with cost and remediation time as criteria. The analytical framework presented allows alternatives to be evaluated in the context of budgetary, social, and political considerations. In general, the longer the remediation takes, the lower the net present value of the process. This is true because of the time value of money and large percentage of the costs attributed to storage or disposal

  4. Mixed waste focus area alternative technologies workshop

    International Nuclear Information System (INIS)

    Borduin, L.C.; Palmer, B.A.; Pendergrass, J.A.

    1995-01-01

    This report documents the Mixed Waste Focus Area (MWFA)-sponsored Alternative Technology Workshop held in Salt Lake City, Utah, from January 24--27, 1995. The primary workshop goal was identifying potential applications for emerging technologies within the Options Analysis Team (OAT) ''wise'' configuration. Consistent with the scope of the OAT analysis, the review was limited to the Mixed Low-Level Waste (MLLW) fraction of DOE's mixed waste inventory. The Los Alamos team prepared workshop materials (databases and compilations) to be used as bases for participant review and recommendations. These materials derived from the Mixed Waste Inventory Report (MWIR) data base (May 1994), the Draft Site Treatment Plan (DSTP) data base, and the OAT treatment facility configuration of December 7, 1994. In reviewing workshop results, the reader should note several caveats regarding data limitations. Link-up of the MWIR and DSTP data bases, while representing the most comprehensive array of mixed waste information available at the time of the workshop, requires additional data to completely characterize all waste streams. A number of changes in waste identification (new and redefined streams) occurred during the interval from compilation of the data base to compilation of the DSTP data base with the end result that precise identification of radiological and contaminant characteristics was not possible for these streams. To a degree, these shortcomings compromise the workshop results; however, the preponderance of waste data was linked adequately, and therefore, these analyses should provide useful insight into potential applications of alternative technologies to DOE MLLW treatment facilities

  5. Elaboration and experimental study of the Borosilicate glass GP 98/12 for the vitrification of the radioactive wastes of KfKarlsruhe Centre (R.F.A.)

    International Nuclear Information System (INIS)

    Ghezal, A.

    1987-09-01

    The transformation into a vitrified block of highly radioactive liquid wastes is actually the best solution for the storage in long run. In West Germany, the research institute in the field of nuclear energy (KfK) has been oriented in this way by developing industrial processes of vitrification and by following studies on the behaviour of the final products. For the fission products, the chosen glasses present good stability characteristics and are used as a first barrier during confinement. Our work, which is part of the research program on radioactive waste vitrification, consists of preparing borosilicate glass GP 98/12 and studying physical and chemical characteristics. We have also contributed to the development and the realization of glass blocks sampling system prepared at pilot scale

  6. 76 FR 13605 - Notice of Availability of Draft Waste Incidental to Reprocessing Evaluation for the Vitrification...

    Science.gov (United States)

    2011-03-14

    ... selected offsite low-level waste disposal facility, either the NNSS Area 5 Radioactive Waste Management... in West Valley, New York, is waste incidental to reprocessing and thus is not high-level radioactive... evaluation pursuant to DOE Manual 435.1-1, Radioactive Waste Management. DOE is consulting with the Nuclear...

  7. Thermal processing systems for TRU mixed waste

    International Nuclear Information System (INIS)

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

    1992-01-01

    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

  8. Mixed waste management at the Hanford Site

    International Nuclear Information System (INIS)

    Roberts, R.J.; Jasen, W.G.

    1991-01-01

    Various types of waste have been generated during the 50-year history of the Hanford Site. Regulatory changes in the last 20 years have provided the emphasis for better management of these wastes. Interpretations of the Atomic Energy Act (AEA) and the Resource Conservation and Recovery Act (RCRA) have led to the definition of a group of wastes called radioactive mixed wastes (RMW). As a result of the radioactive and hazardous properties of these wastes, special projects have been initiated for the management of RMW. This paper addresses the management of solid RMW. The management of bulk liquid RMW will not be described. 7 refs., 4 figs

  9. The mixed waste landfill integrated demonstration

    International Nuclear Information System (INIS)

    Burford, T.D.; Williams, C.V.

    1994-01-01

    The Mixed Waste Landfill Integrated Demonstration (MWLID) focuses on ''in-situ'' characterization, monitoring, remediation, and containment of landfills in arid environments that contain hazardous and mixed waste. The MWLID mission is to assess, demonstrate, and transfer technologies and systems that lead to faster, better, cheaper, and safer cleanup. Most important, the demonstrated technologies will be evaluated against the baseline of conventional technologies and systems. The comparison will include the cost, efficiency, risk, and feasibility of using these innovative technologies at other sites

  10. Certification Plan, Radioactive Mixed Waste Hazardous Waste Handling Facility

    International Nuclear Information System (INIS)

    Albert, R.

    1992-01-01

    The purpose of this plan is to describe the organization and methodology for the certification of radioactive mixed waste (RMW) handled in the Hazardous Waste Handling Facility at Lawrence Berkeley Laboratory (LBL). RMW is low-level radioactive waste (LLW) or transuranic (TRU) waste that is co-contaminated with dangerous waste as defined in the Westinghouse Hanford Company (WHC) Solid Waste Acceptance Criteria (WAC) and the Washington State Dangerous Waste Regulations, 173-303-040 (18). This waste is to be transferred to the Hanford Site Central Waste Complex and Burial Grounds in Hanford, Washington. This plan incorporates the applicable elements of waste reduction, which include both up-front minimization and end-product treatment to reduce the volume and toxicity of the waste; segregation of the waste as it applies to certification; an executive summary of the Waste Management Quality Assurance Implementing Management Plan (QAIMP) for the HWHF (Section 4); and a list of the current and planned implementing procedures used in waste certification

  11. Process performance of the pilot-scale in situ vitrification of a simulated waste disposal site at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Carter, J.G.; Koegler, S.S.; Bates, S.O.

    1988-06-01

    Process feasibility studies have been successfully performed on three developmental scales to determine the potential for applying in situ vitrification to intermediate-level (low-level) waste placed in seepage pits and trenches at Oak Ridge National Laboratory (ORNL). In the laboratory, testing was performed in crucibles containing a mixture of 50% ORNL soil and 50% limestone. In an engineering-scale test at Pacific Northwest Laboratory a /1/12/-scale simulation of an ORNL waste trench was constructed and vitrified, resulting in a waste product containing soil and limestone concentrations of 68 wt % and 32 wt %, respectively. In the pilot-scale test a /3/8/-scale simulation of the same trench was constructed and vitrified at ORNL, resulting in soil and limestone concentrations of 80% and 20%, respectively, in the waste product. Results of the three scales of testing indicate that the ORNL intermediate-level (low-level) waste sites can be successfully processed by in situ vitrification; the waste form will retain significant quantities of the cesium and strontium. Because cesium-137 and strontium-90 are the major components of the radionuclide inventory in the ORNL seepage pits and trenches, final field process decontamination factors (i.e., losses to the off-gas system relative to the waste inventory) of 1.0 E + 4 are desired to minimize activity buildup in the off-gas system. 17 refs., 34 figs., 13 tabs

  12. Development of the vitrification compositional envelope to support complex-wide application of MAWS technology

    International Nuclear Information System (INIS)

    Mazer, J.J.; Muller, I.S.; Gan, H.; Buechele, A.C.; Lai, S.T.; Pegg, I.L.

    1996-09-01

    This report presents the results from a study of the application of the Minimum Additive Waste Stabilization (MAWS) approach using vitrification as a treatment technology to a variety of waste streams across the DOE complex. This work has involved both experimental vitrification work using actual mixed wastes and surrogate waste streams from several DOE sites (Hanford, Idaho, and Oak Ridge) as well as the development of a computer-based, integrated glass property-composition database. The long-term objective is that this data base will assist glass formulation studies with single waste streams or combinations of waste streams subject to a variety of user-imposed constraints including waste stream usage priorities, process related constraints (e.g., melt viscosity, electrical conductivity, etc.), and waste form performance related constraints (e.g., TCLP and PCT leaching results). 79 refs., 143 figs., 65 tabs

  13. Conversion of three mixed-waste streams

    International Nuclear Information System (INIS)

    Harmer, D.E.; Porter, D.L.; Conley, C.W.

    1990-01-01

    At the present time, commercial mixed waste (containing both radioactive and hazardous components) is not handled by any disposal site in this country. Thus, a generator of such material is faced with the prospect of separating or altering the nature of the waste components. A chemical or physical separation may be possible. However, if separation fails there remains the opportunity of chemically transforming the hazardous ingredients to non-hazardous substances, allowing disposal at an existing radioactive burial site. Finally, chemical or physical stabilization can be used as a tool to achieve an acceptable waste form lacking the characteristics of mixed waste. A practical application of these principles has been made in the case of certain mixed waste streams at Aerojet Ordnance Tennessee. Three different streams were involved: (1) lead and lead oxide contaminated with uranium, (2) mixed chloride salts including barium chloride, contaminated with uranium, and (3) bricks impregnated with the barium salt mixture. This paper summarizes the approach of this mixed-waste problem, the laboratory solutions found, and the intended field remediations to be followed. Mixture (1), above, was successfully converted to a vitreous insoluble form. Mixture (2) was separated into radioactive and non-radioactive streams, and the hazardous characteristics of the latter altered chemically. Mixture (3) was treated to an extraction process, after which the extractant could be treated by the methods of Mixture (2). Field application of these methods is scheduled in the near future

  14. Pump Jet Mixing and Pipeline Transfer Assessment for High-Activity Radioactive Wastes in Hanford Tank 241-AZ-102

    Energy Technology Data Exchange (ETDEWEB)

    Y Onishi; KP Recknagle; BE Wells

    2000-08-09

    The authors evaluated how well two 300-hp mixer pumps would mix solid and liquid radioactive wastes stored in Hanford double-shell Tank 241-AZ-102 (AZ-102) and confirmed the adequacy of a three-inch (7.6-cm) pipeline system to transfer the resulting mixed waste slurry to the AP Tank Farm and a planned waste treatment (vitrification) plant on the Hanford Site. Tank AZ-102 contains 854,000 gallons (3,230 m{sup 3}) of supernatant liquid and 95,000 gallons (360 m{sup 3}) of sludge made up of aging waste (or neutralized current acid waste). The study comprises three assessments: waste chemistry, pump jet mixing, and pipeline transfer. The waste chemical modeling assessment indicates that the sludge, consisting of the solids and interstitial solution, and the supernatant liquid are basically in an equilibrium condition. Thus, pump jet mixing would not cause much solids precipitation and dissolution, only 1.5% or less of the total AZ-102 sludge. The pump jet mixing modeling indicates that two 300-hp mixer pumps would mobilize up to about 23 ft (7.0 m) of the sludge nearest the pump but would not erode the waste within seven inches (0.18 m) of the tank bottom. This results in about half of the sludge being uniformly mixed in the tank and the other half being unmixed (not eroded) at the tank bottom.

  15. Pump Jet Mixing and Pipeline Transfer Assessment for High-Activity Radioactive Wastes in Hanford Tank 241-AZ-102

    International Nuclear Information System (INIS)

    Onishi, Y.; Recknagle, K.P.; Wells, B.E.

    2000-01-01

    The authors evaluated how well two 300-hp mixer pumps would mix solid and liquid radioactive wastes stored in Hanford double-shell Tank 241-AZ-102 (AZ-102) and confirmed the adequacy of a three-inch (7.6-cm) pipeline system to transfer the resulting mixed waste slurry to the AP Tank Farm and a planned waste treatment (vitrification) plant on the Hanford Site. Tank AZ-102 contains 854,000 gallons (3,230 m 3 ) of supernatant liquid and 95,000 gallons (360 m 3 ) of sludge made up of aging waste (or neutralized current acid waste). The study comprises three assessments: waste chemistry, pump jet mixing, and pipeline transfer. The waste chemical modeling assessment indicates that the sludge, consisting of the solids and interstitial solution, and the supernatant liquid are basically in an equilibrium condition. Thus, pump jet mixing would not cause much solids precipitation and dissolution, only 1.5% or less of the total AZ-102 sludge. The pump jet mixing modeling indicates that two 300-hp mixer pumps would mobilize up to about 23 ft (7.0 m) of the sludge nearest the pump but would not erode the waste within seven inches (0.18 m) of the tank bottom. This results in about half of the sludge being uniformly mixed in the tank and the other half being unmixed (not eroded) at the tank bottom

  16. Status of mixed-waste regulation

    International Nuclear Information System (INIS)

    Bahadur, S.

    1988-01-01

    Mixed waste is waste containing radionuclides regulated by the US Nuclear Regulatory Commission (NRC) under the Atomic Energy Act, as well as hazardous waste materials regulated by the Environmental Protection Agency (EPA) under the Resource Conservation and Recovery Act (RCRA). This has led to a situation of dual regulation in which both NRC and EPA regulate the same waste under requirements that at times appear conflicting. The NRC has been working with the EPA to resolve the issues associated with the dual regulation of mixed waste. Discussions between the two agencies indicate that dual regulation of mixed wastes appears technically achievable, although the procedures may be complex and burdensome to the regulated community. The staffs of both agencies have been coordinating their efforts to minimize the burden of dual regulation on state agencies and the industry. Three major issues were identified as sources of potential regulatory conflict: (a) definition and identification of mixed waste, (b) siting guidelines for disposal facilities, and (c) design concepts for disposal units

  17. Bioprocessing of a stored mixed liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Wolfram, J.H.; Rogers, R.D. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Finney, R. [Mound Applied Technologies, Miamisburg, OH (United States)] [and others

    1995-12-31

    This paper describes the development and results of a demonstration for a continuous bioprocess for mixed waste treatment. A key element of the process is an unique microbial strain which tolerates high levels of aromatic solvents and surfactants. This microorganism is the biocatalysis of the continuous flow system designed for the processing of stored liquid scintillation wastes. During the past year a process demonstration has been conducted on commercial formulation of liquid scintillation cocktails (LSC). Based on data obtained from this demonstration, the Ohio EPA granted the Mound Applied Technologies Lab a treatability permit allowing the limited processing of actual mixed waste. Since August 1994, the system has been successfully processing stored, {open_quotes}hot{close_quotes} LSC waste. The initial LSC waste fed into the system contained 11% pseudocumene and detectable quantities of plutonium. Another treated waste stream contained pseudocumene and tritium. Data from this initial work shows that the hazardous organic solvent, and pseudocumene have been removed due to processing, leaving the aqueous low level radioactive waste. Results to date have shown that living cells are not affected by the dissolved plutonium and that 95% of the plutonium was sorbed to the biomass. This paper discusses the bioprocess, rates of processing, effluent, and the implications of bioprocessing for mixed waste management.

  18. Co-disposal of mixed waste materials

    International Nuclear Information System (INIS)

    Phillips, S.J.; Alexander, R.G.; Crane, P.J.; England, J.L.; Kemp, C.J.; Stewart, W.E.

    1993-08-01

    Co-disposal of process waste streams with hazardous and radioactive materials in landfills results in large, use-efficiencies waste minimization and considerable cost savings. Wasterock, produced from nuclear and chemical process waste streams, is segregated, treated, tested to ensure regulatory compliance, and then is placed in mixed waste landfills, burial trenches, or existing environmental restoration sites. Large geotechnical unit operations are used to pretreat, stabilize, transport, and emplace wasterock into landfill or equivalent subsurface structures. Prototype system components currently are being developed for demonstration of co-disposal

  19. Commercialization project of Ulchin vitrification

    International Nuclear Information System (INIS)

    Jo, Hyun-Jun; Kim, Cheon-Woo; Hwang, Tae-Won

    2011-01-01

    The Ulchin Vitrification Facility (UVF), to be used for the vitirification of low- and intermediate-level radioactive waste (LILW) generated by nuclear power plants (NPPs), is the world's first commercial facility using Cold Crucible Induction Melter (CCIM) technology. The construction of the facility was begun in 2005 and was completed in 2007. From December 2007 to September 2009, all key performance tests, such as the system functional test, the cold test, the hot test, and the real waste test, were successfully carried out. The UVF commenced commercial operation in October 2009 for the vitrification of radioactive waste. (author)

  20. Mixed Low-Level Radioactive Waste (MLLW) Primer

    International Nuclear Information System (INIS)

    Schwinkendorf, W.E.

    1999-01-01

    This document presents a general overview of mixed low-level waste, including the regulatory definitions and drivers, the manner in which the various kinds of mixed waste are regulated, and a discussion of the waste treatment options

  1. Mixed Low-Level Radioactive Waste (MLLW) Primer

    Energy Technology Data Exchange (ETDEWEB)

    W. E. Schwinkendorf

    1999-04-01

    This document presents a general overview of mixed low-level waste, including the regulatory definitions and drivers, the manner in which the various kinds of mixed waste are regulated, and a discussion of the waste treatment options.

  2. Vitrified waste form performance modeling applied to the treatment and disposal of mixed-waste sludge at the Savannah River Site

    International Nuclear Information System (INIS)

    Whited, A.R.; Fjeld, R.A.

    1998-01-01

    Vitrification, the conversion of source components into a solid amorphous glass matrix, has emerged as a viable treatment technology for low-level radioactive waste and mixed waste. To dispose of vitrified low-level waste at US Department of Energy facilities, site-specific radiological performance assessments must be conducted to demonstrate that waste glass satisfies performance objectives for environmental protection. More than 2,500 m 3 of F0006-listed low-level mixed-waste sludge stored in the Reactor Materials Department (M-Area) at the Savannah River Site (SRS) is scheduled for vitrification. This study evaluates the feasibility of on-site disposal of vitrified M-Area waste at SRS. Laboratory leaching tests that accelerate the glass corrosion process are currently the best indicators of vitrified waste form durability. A method to incorporate laboratory leaching data into performance assessments is presented. A screening-level performance assessments code is used to model trench disposal of M-Area waste glass. The allowable leach rate for vitrified M-Area waste is determined based on both a maximum radiological dose equivalent of 4 mrem/yr for the drinking water pathway and a maximum uranium concentration of 20 microg/ell in groundwater. The allowable leach rate is compared with published long-term leaching data for a wide range of waste glass compositions and test conditions. This analysis demonstrates that trench disposal of the waste glass is likely to meet applicable performance objectives if the glass is of above average durability compared with the reference set of glasses

  3. Vitrification: An alternative to minimize environmental impact caused by leather industry wastes

    International Nuclear Information System (INIS)

    Basegio, T.; Beck Leao, A.P.; Bernardes, A.M.; Bergmann, C.P.

    2009-01-01

    The main purpose of this work was the investigation of the immobilization of chromium ion present in the ash from the incineration of chromium-tanned leather shavings (ACS) by means of vitrified ceramic bodies. To achieve the immobilization samples were initially produced adding soda-lime glass to ACS. After that, new formulations were produced with the addition of pure oxides, TiO 2 and MgO, to the soda-lime glass and ACS sample. These samples were conformed by pressing and fired in an eletrical furnace, at temperatures of 750, 800, 950 and 1000 o C. The ceramic bodies were evaluated with respect to their physical properties and mineralogical composition by X-ray diffraction and energy dispersive spectroscopy (EDS) mapping. The chromium immobilization was characterized by leaching tests according to Brazilian Regulations (NBR 10005). The results confirmed the immobilization of the chromium ion within the allowed limits of Brazilian Regulations, NBR 10004 (5 mg/L), with the use of glass and vitrification/densification agents such as titanium and magnesium oxides.

  4. Performance of High Temperature Filter System for Radioactive Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Park, Seung Chul; Hwang, Tae Won; Shin, Sang Won; Ha, Jong Hyun; Kim, Hey Suk; Park, So Jin

    2004-01-01

    Important operation parameters and performance of a high temperature ceramic candle filter system were evaluated through a series of demonstration tests at a pilot-scale vitrification plant. At the initial period of each test, due to the growth of dust cake on the surface of ceramic candles, the pressure drop across the filter media increased sharply. After that it became stable to a certain range and varied continuously proportion to the face velocity of off-gas. On the contrary, at the initial period of each test, the permeability of filter element decreased rapidly and then it became stable. Back flushing of the filter system was effective under the back flushing air pressure range of 3∼5 bar. Based on the dust concentrations measured by iso-kinetic dust sampling at the inlet and outlet point of HTF, the dust collection efficiency of HTF evaluated. The result met the designed performance value of 99.9%. During the demonstration tests including a hundred hour long test, no specific failure or problem affecting the performance of HTF system were observed.

  5. Small-scale, joule-heated melting of Savannah River Plant waste glass. I. Factors affecting large-scale vitrification tests

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Chismar, P.H.

    1979-10-01

    A promising method of immobilizing SRP radioactive waste solids is incorporation in borosilicate glass. In the reference vitrification process, called joule-heated melting, a mixture of glass frit and calcined waste is heated by passage of an electric current. Two problems observed in large-scale tests are foaming and formation of an insoluble slag. A small joule-heated melter was designed and built to study problems such as these. This report describes the melter, identifies factors involved in foaming and slag formation, and proposes ways to overcome these problems

  6. Formulation Efforts for Direct Vitrification of INEEL Blend Calcine Waste Simulate: Fiscal Year 2000

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Vienna, John D.; Peeler, David K.; Reamer, I. A.

    2001-03-30

    This report documents the results of glass formulation efforts for Idaho National Engineering and Environmental Laboratory (INEEL) high level waste (HWL) calcine. Two waste compositions were used during testing. Testing started by using the Run 78 calcine composition and switched to simulated Blend calcine composition when it became available. The goal of the glass formulation efforts was to develop a frit composition that will accept higher waste loading that satisfies the glass processing and product acceptance constraints. 1. Melting temperature of 1125 ? 25?C 2. Viscosity between 2 and 10 Pa?s at the melting temperature 3. Liquidus temperature at least 100?C below the melting temperature 4. Normalized release of B, Li and Na each below 1 g/m2 (per ASTM C 1285-97) Glass formulation efforts tested several frit compositions with variable waste loadings of Run 78 calcine waste simulant. Frit 107 was selected as the primary candidate for processing since it met all process and performance criteria up to 45 mass% waste loading. When the simulated Blend calcine waste composition became available Frits 107 and 108 compositions were retested and again Frit 107 remained the primary candidate. However, both frits suffered a decrease in waste loading when switching from the Run 78 calcine to simulated Blend calcine waste composition. This was due to increase concentrations of both F and Al2O3 along with a decrease in CaO and Na2O in the simulate Blend calcine waste all of which have strong impacts on the glass properties that limit waste loading of this type of waste.

  7. Preliminary estimates of cost savings for defense high level waste vitrification options

    International Nuclear Information System (INIS)

    Merrill, R.A.; Chapman, C.C.

    1993-09-01

    The potential for realizing cost savings in the disposal of defense high-level waste through process and design modificatins has been considered. Proposed modifications range from simple changes in the canister design to development of an advanced melter capable of processing glass with a higher waste loading. Preliminary calculations estimate the total disposal cost (not including capital or operating costs) for defense high-level waste to be about $7.9 billion dollars for the reference conditions described in this paper, while projected savings resulting from the proposed process and design changes could reduce the disposal cost of defense high-level waste by up to $5.2 billion

  8. Requirements for permitting a mixed waste incinerator

    International Nuclear Information System (INIS)

    Trichon, M.; Feldman, J.; Serne, J.C.

    1990-01-01

    The consideration, design, selection and operation of any incinerator depends primarily on characteristic quality (ultimate and proximate analyses) and quantity to the waste to be incinerated. In the case of burning any combination of mixed hazardous, biomedical and radioactive low level waste, specific federal and generic state environmental regulatory requirements are outlined. Combustion chamber temperature and waste residence time requirements will provide the rest of the envelope for consideration. Performance requirements must be balanced between the effects of time and temperature on destruction of the organic waste and the vaporization and possible emission of the inorganic waste components (e.g., toxic metals, radioactive inorganics) as operating conditions and emission levels will be set in state and federal regulatory permits. To this end the complete characterization of the subject waste stream must be determined if an accurate assessment of incineration effectiveness and impact are to be performed

  9. Optimization of glass composition for the vitrification of nuclear waste at the Savannah River Plant

    International Nuclear Information System (INIS)

    Soper, P.D.; Roberts, G.J.; Lightner, L.F.; Walker, D.D.; Plodinec, M.J.

    1982-01-01

    Waste glasses of different compositions were compared in terms of leachability, viscosity, liquidus temperature, and coefficient of expansion. The compositions of the glasses were determined by statistical optimization. Waste glass of the optimized composition is more durable than the current reference composition but can still be processed at low temperature

  10. Savannah River Plant defense waste vitrification studies during FY 1982. Summary report

    International Nuclear Information System (INIS)

    Ethridge, L.J.

    1983-10-01

    Five major melter runs were completed during FY 1982 on the Pilot-Scale Ceramic Melter (PSCM). Over 41,000 L of feed were processed by the PSCM, producing approx. 21,000 kg of glass. The design basis reference capacity of approx. 39 kg/h-m 2 was met or exceeded in all the melter runs. Off-gas characterization was emphasized during this fiscal year. Entrainment of feed material is the largest contributor to the mass of particulate leaving the melter, averaging 0.2 wt% of the incoming feed on an oxide basis. This is a DF of approx. 500. This mass does show an enrichment of some of the volatile and semivolatile components. Higher losses of cesium, tellurium, and cadmium occurred with formate feed. The Experimental Ceramic Melter (ECM) was used this year to study the application of two techniques to increase melting rates in ceramic melters. The first was the use of an air sparger to forcibly agitate the glass in the melter to improve the heat transfer. The air-sparger agitation increased the throughput capacity of the ECM, but did not seem to affect melting efficiency. The second technique for increasing melter rates tested on the ECM was the use of microwave boosting. While significant improvement was noted in the vitrification rates, two problems were encountered: coating of the isolation window and heating of the refractory lining of the ECM lid. The buildup of fine dust on the window caused arcing between the coating and the waveguide. This arcing damages the window and waveguide and causes instability in the microwave power supply. Four techniques were investigated to solve the problem. These techniques were of limited success and await further testing. 33 figures, 58 tables

  11. The Hazardous Waste/Mixed Waste Disposal Facility

    International Nuclear Information System (INIS)

    Bailey, L.L.

    1991-01-01

    The Hazardous Waste/Mixed Waste Disposal Facility (HW/MWDF) will provide permanent Resource Conservation and Recovery Act (RCRA) permitted storage, treatment, and disposal for hazardous and mixed waste generated at the Department of Energy's (DOE) Savannah River Site (SRS) that cannot be disposed of in existing or planned SRS facilities. Final design is complete for Phase I of the project, the Disposal Vaults. The Vaults will provide RCRA permitted, above-grade disposal capacity for treated hazardous and mixed waste generated at the SRS. The RCRA Part B Permit application was submitted upon approval of the Permit application, the first Disposal Vault is scheduled to be operational in mid 1994. The technical baseline has been established for Phase II, the Treatment Building, and preliminary design work has been performed. The Treatment Building will provide RCRA permitted treatment processes to handle a variety of hazardous and mixed waste generated at SRS in preparation for disposal. The processes will treat wastes for disposal in accordance with the Environmental Protection Agency's (EPA's) Land Disposal Restrictions (LDR). A RCRA Part B Permit application has not yet been submitted to SCDHEC for this phase of the project. The Treatment Building is currently scheduled to be operational in late 1996

  12. Hazardous and mixed waste transportation program

    International Nuclear Information System (INIS)

    Hohnstreiter, G.F.; Glass, R.E.; McAllaster, M.E.; Nigrey, P.J.; Trennel, A.J.; Yoshimura, H.R.

    1993-01-01

    Sandia National Laboratories (SNL) has developed a program to address the packaging needs associated with the transport of hazardous and mixed waste during the United States' Department of Energy (DOE) remediation efforts. The program addresses the technology needs associated with the transport of materials which have components that are radioactive and chemically hazardous. The mixed waste transportation activities focus on on-site specific applications of technology to the transport of hazardous and mixed wastes. These activities were identified at a series of DOE-sponsored workshops. These activities will be composed of the following: (1) packaging concepts, (2) chemical compatibility studies, and (3) systems studies. This paper will address activities in each of these areas. (J.P.N.)

  13. Hazardous and Mixed Waste Transportation Program

    International Nuclear Information System (INIS)

    Hohnstreiter, G.F.; Glass, R.E.; McAllaster, M.E.; Nigrey, P.J.; Trennel, A.J.; Yoshimura, H.R.

    1991-01-01

    Sandia National Laboratories (SNL) has developed a program to address the packaging needs associated with the transport of hazardous and mixed waste during the United States' Department of Energy (DOE) remediation efforts. The program addresses the technology needs associated with the transport of materials which have components that are radioactive and chemically hazardous. The mixed waste transportation activities focus on on-site specific applications of technology to the transport of hazardous and mixed wastes. These activities were identified at a series of DOE-sponsored workshops. These activities will be composed of the following: (1) packaging concepts, (2) chemical compatibility studies, and (3) systems studies. This paper will address activities in each of these areas

  14. Vitrification of radioactive high-level waste by spray calcination and in-can melting

    Science.gov (United States)

    Hanson, M. S.; Bjorklund, W. J.

    1980-07-01

    After several nonradioactive test runs, radioactive waste from the processing of 1.5 t of spent, light water reactor fuel was successfully concentrated, dried and converted to a vitreous product. A total of 97 L of waste glass (in two stainless steel canisters) was produced. The spray calcination process coupled to the in-can melting process, as developed at Pacific Northwest Labortory, was used to vitrify the waste. An effluent system consisting of a variety of condensation of scrubbing steps more than adequately decontaminated the process off gas before it was released to the atmosphere.

  15. Mixed waste treatment capabilities at Envirocare

    International Nuclear Information System (INIS)

    Rafati, A.

    1994-01-01

    This presentation gives an overview of the business achievements and presents a corporate summary for the whole handling company Envirocare located in Clive, Utah. This company operates a permitted low-level radioactive and mixed waste facility which handles waste from the United States Department of Energy, Environmental Protection Agency, Department of Defense, and Fortune 500 companies. A description of business services and treatment capabilities is presented

  16. Mixed Waste Integrated Program emerging technology development

    International Nuclear Information System (INIS)

    Berry, J.B.; Hart, P.W.

    1994-01-01

    The US Department of Energy (DOE) is responsible for the management and treatment of its mixed low-level wastes (MLLW). MLLW are regulated under both the Resource Conservation and Recovery Act and various DOE orders. Over the next 5 years, DOE will manage over 1.2 m 3 of MLLW and mixed transuranic (MTRU) wastes. In order to successfully manage and treat these mixed wastes, DOE must adapt and develop characterization, treatment, and disposal technologies which will meet performance criteria, regulatory approvals, and public acceptance. Although technology to treat MLLW is not currently available without modification, DOE is committed to developing such treatment technologies and demonstrating them at the field scale by FY 1997. The Office of Research and Development's Mixed Waste Integrated Program (MWIP) within the DOE Office of Environmental Management (EM), OfFice of Technology Development, is responsible for the development and demonstration of such technologies for MLLW and MTRU wastes. MWIP advocates and sponsors expedited technology development and demonstrations for the treatment of MLLW

  17. Molten salt destruction process for mixed wastes

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Wilder, J.G.; Karlsen, C.E.

    1993-04-01

    We are developing an advanced two-stage process for the treatment of mixed wastes, which contain both hazardous and radioactive components. The wastes, together with an oxidant gas, such as air, are injected into a bed of molten salt comprising a mixture of sodium-, potassium-, and lithium-carbonates, with a melting point of about 580 degree C. The organic constituents of the mixed waste are destroyed through the combined effect of pyrolysis and oxidation. Heteroatoms. such as chlorine, in the mixed waste form stable salts, such as sodium chloride, and are retained in the melt. The radioactive actinides in the mixed waste are also retained in the melt because of the combined action of wetting and partial dissolution. The original process, consists of a one-stage unit, operated at 900--1000 degree C. The advanced two-stage process has two stages, one for pyrolysis and one for oxidation. The pyrolysis stage is designed to operate at 700 degree C. The oxidation stage can be operated at a higher temperature, if necessary

  18. Mixed Waste Integrated Program emerging technology development

    Energy Technology Data Exchange (ETDEWEB)

    Berry, J.B. [Oak Ridge National Lab., TN (United States); Hart, P.W. [USDOE, Washington, DC (United States)

    1994-06-01

    The US Department of Energy (DOE) is responsible for the management and treatment of its mixed low-level wastes (MLLW). MLLW are regulated under both the Resource Conservation and Recovery Act and various DOE orders. Over the next 5 years, DOE will manage over 1.2 m{sup 3} of MLLW and mixed transuranic (MTRU) wastes. In order to successfully manage and treat these mixed wastes, DOE must adapt and develop characterization, treatment, and disposal technologies which will meet performance criteria, regulatory approvals, and public acceptance. Although technology to treat MLLW is not currently available without modification, DOE is committed to developing such treatment technologies and demonstrating them at the field scale by FY 1997. The Office of Research and Development`s Mixed Waste Integrated Program (MWIP) within the DOE Office of Environmental Management (EM), OfFice of Technology Development, is responsible for the development and demonstration of such technologies for MLLW and MTRU wastes. MWIP advocates and sponsors expedited technology development and demonstrations for the treatment of MLLW.

  19. Mixed waste paper to ethanol fuel

    Energy Technology Data Exchange (ETDEWEB)

    1991-01-01

    The objectives of this study were to evaluate the use of mixed waste paper for the production of ethanol fuels and to review the available conversion technologies, and assess developmental status, current and future cost of production and economics, and the market potential. This report is based on the results of literature reviews, telephone conversations, and interviews. Mixed waste paper samples from residential and commercial recycling programs and pulp mill sludge provided by Weyerhauser were analyzed to determine the potential ethanol yields. The markets for ethanol fuel and the economics of converting paper into ethanol were investigated.

  20. Mixed radioactive and chemotoxic wastes (RMW)

    International Nuclear Information System (INIS)

    Dejonghe, I.P.

    1991-01-01

    During the first decades of development of nuclear energy, organizations involved in the management of nuclear wastes had their attention focused essentially on radioactive components. The impression may have prevailed that, considering the severe restrictions on radioactive materials, the protection measured applied for radioactive components of wastes would be more than adequate to cope with potential hazards from non radioactive components associated with radioactive wastes. More recently it was acknowledged that such interpretation is not necessarily justified in all cases since certain radioactive wastes also contain non-negligible amounts of heavy metals or hazardous organic components which, either, do not decay, or are subject to completely different decay (decomposition) mechanisms. The main purposes of the present study are to analyze whether mixed radioactive wastes are likely to occur in Europe and in what form, whether one needs a basis for integration for evaluating various forms of toxicity and by which practical interventions possible problems can be avoided or at least reduced. (au)

  1. Mixed low-level waste form evaluation

    International Nuclear Information System (INIS)

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

    1997-01-01

    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

  2. Small-scale integrated demonstration of high-level radioactive waste processing and vitrification using actual SRP waste

    International Nuclear Information System (INIS)

    Ferguson, R.B.; Woolsey, G.B.; Galloway, R.M.; Baumgarten, P.M.; Eibling, R.E.

    1980-01-01

    Experiments have been made to demonstrate the feasibility of immobilizing SRP high-level waste in borosilicate glass. Results to date are encouraging. Equipment performance and processing characteristics for solidifying small batches of actual SRP waste have agreed well with previous experience with small- and large-scale tests synthetic waste, and with theoretical predictions

  3. Letter report: Pre-conceptual design study for a pilot-scale Non-Radioactive Low-Level Waste Vitrification Facility

    International Nuclear Information System (INIS)

    Thompson, R.A.; Morrissey, M.F.

    1996-03-01

    This report presents a pre-conceptual design study for a Non-Radioactive Low-Level Waste, Pilot-Scale Vitrification System. This pilot plant would support the development of a full-scale LLW Vitrification Facility and would ensure that the full-scale facility can meet its programmatic objectives. Use of the pilot facility will allow verification of process flowsheets, provide data for ensuring product quality, assist in scaling to full scale, and support full-scale start-up. The facility will vitrify simulated non-radioactive LLW in a manner functionally prototypic to the full-scale facility. This pre-conceptual design study does not fully define the LLW Pilot-Scale Vitrification System; rather, it estimates the funding required to build such a facility. This study includes identifying all equipment necessary. to prepare feed, deliver it into the melter, convert the feed to glass, prepare emissions for atmospheric release, and discharge and handle the glass. The conceived pilot facility includes support services and a structure to contain process equipment

  4. Chloride removal from vitrification offgas

    Energy Technology Data Exchange (ETDEWEB)

    Slaathaug, E.J. [Westinghouse Hanford Co., Richland, WA (United States)

    1995-06-01

    This study identified and investigated techniques of selectively purging chlorides from the low-level waste (LLW) vitrification process with the purge stream acceptable for burial on the Hanford Site. Chlorides will be present in high concentration in several individual feeds to the LLW Vitrification Plant. The chlorides are highly volatile in combustion type melters and are readily absorbed by wet scrubbing of the melter offgas. The Tank Waste Remediation System (TWRS) process flow sheets show that the resulting chloride rich scrub solution is recycled back to the melter. The chlorides must be purged from the recycle loop to prevent the buildup of excessively high chloride concentrations.

  5. Chloride removal from vitrification offgas

    International Nuclear Information System (INIS)

    Slaathaug, E.J.

    1995-01-01

    This study identified and investigated techniques of selectively purging chlorides from the low-level waste (LLW) vitrification process with the purge stream acceptable for burial on the Hanford Site. Chlorides will be present in high concentration in several individual feeds to the LLW Vitrification Plant. The chlorides are highly volatile in combustion type melters and are readily absorbed by wet scrubbing of the melter offgas. The Tank Waste Remediation System (TWRS) process flow sheets show that the resulting chloride rich scrub solution is recycled back to the melter. The chlorides must be purged from the recycle loop to prevent the buildup of excessively high chloride concentrations

  6. AN ALTERNATIVE HOST MATRIX BASED ON IRON PHOSPHATE GLASSES FOR THE VITRIFICATION OF SPECIALIZED WASTE FORMS

    International Nuclear Information System (INIS)

    Day, Delbert D.

    2000-01-01

    As mentioned above, the overall goal of this research project was to collect the scientific information essential to develop iron phosphate glass based nuclear wasteforms. The specific objectives of the project were: (1) Investigate the structure of binary iron phosphate glasses and it's dependence on the composition and melting atmosphere: Understand atomic arrangements and nature of the bonding. Establish structure-property relationships. Determine the compositions and melting conditions which optimize the critical properties of the base glass. (2) Understand the structure of iron phosphate wasteforms and it's dependence on the composition and melting atmosphere: Investigate how the waste elements are bonded and coordinated within the glass structure. Establish structure-property relationships for the waste glasses. Determine the compositions and melting atmosphere for which the critical properties of the waste forms would be optimum. (3) Determine the role(s) played by the valence states of iron ions and it's dependence on the composition and melting atmosphere: Understand the different roles of iron(II) and iron(III) ions in determining the critical properties of the base glass and the waste forms. Investigate how the iron valence and its significance depend on the composition and melting atmosphere. (4) Investigate glass forming and crystallization processes of the iron phosphate glasses and their waste forms: Understand the dependence of the glass forming and crystallization characteristics on overall glass composition and valence states of iron ions. Identify the products of devitrification and investigate the critical properties of these crystalline compounds which may adversely affect the chemical and physical properties of the waste forms

  7. Vitrification and Product Testing of AW-101 and AN-107 Pretreated Waste

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L.; Greenwood, Lawrence R.; Piepel, Gregory F.; Schweiger, Michael J.; Smith, Harry D.; Urie, Michael W.; Wagner, Jerome J.

    2000-10-31

    The primary objective for vitrifying the LAW samples is to generate glass products for subsequent product testing. The work presented in this report is divided into 6 work elements: 1) Glass Fabrication, 2) Chemical Composition, 3) Radiochemical Composition, 4) Crystalline and Non-crystalline Phase Determination, and 5) Release Rate (Modified PCT). These work elements will help demonstrate the RPP-WTP projects ability to satisfy the product requirements concerning, chemical and radionuclide reporting, waste loading, identification and quantification of crystalline and non-crystalline phases, and waste form leachability. VOA, SVOA, dioxins, furans, PCBs, and total cyanide analyses will be reported in as separate document (WTP-RPT-005).

  8. Sulfur polymer cement for macroencapsulation of mixed waste debris

    International Nuclear Information System (INIS)

    Mattus, C.H.

    1998-01-01

    In FY 1997, the US DOE Mixed Waste Focus Area (MWFA) sponsored a demonstration of the macroencapsulation of mixed waste debris using sulfur polymer cement (SPC). Two mixed wastes were tested--a D006 waste comprised of sheets of cadmium and a D008/D009 waste comprised of lead pipes and joints contaminated with mercury. The demonstration was successful in rendering these wastes compliant with Land Disposal Restrictions (LDR), thereby eliminating one Mixed Waste Inventory Report (MWIR) waste stream from the national inventory

  9. Methodology to remediate a mixed waste site

    Energy Technology Data Exchange (ETDEWEB)

    Berry, J.B.

    1994-08-01

    In response to the need for a comprehensive and consistent approach to the complex issue of mixed waste management, a generalized methodology for remediation of a mixed waste site has been developed. The methodology is based on requirements set forth in the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and the Resource Conservation and Recovery Act (RCRA) and incorporates ``lessons learned`` from process design, remediation methodologies, and remediation projects. The methodology is applied to the treatment of 32,000 drums of mixed waste sludge at the Oak Ridge K-25 Site. Process technology options are developed and evaluated, first with regard to meeting system requirements and then with regard to CERCLA performance criteria. The following process technology options are investigated: (1) no action, (2) separation of hazardous and radioactive species, (3) dewatering, (4) drying, and (5) solidification/stabilization. The first two options were eliminated from detailed consideration because they did not meet the system requirements. A quantitative evaluation clearly showed that, based on system constraints and project objectives, either dewatering or drying the mixed waste sludge was superior to the solidification/stabilization process option. The ultimate choice between the drying and the dewatering options will be made on the basis of a technical evaluation of the relative merits of proposals submitted by potential subcontractors.

  10. Methodology to remediate a mixed waste site

    International Nuclear Information System (INIS)

    Berry, J.B.

    1994-08-01

    In response to the need for a comprehensive and consistent approach to the complex issue of mixed waste management, a generalized methodology for remediation of a mixed waste site has been developed. The methodology is based on requirements set forth in the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and the Resource Conservation and Recovery Act (RCRA) and incorporates ''lessons learned'' from process design, remediation methodologies, and remediation projects. The methodology is applied to the treatment of 32,000 drums of mixed waste sludge at the Oak Ridge K-25 Site. Process technology options are developed and evaluated, first with regard to meeting system requirements and then with regard to CERCLA performance criteria. The following process technology options are investigated: (1) no action, (2) separation of hazardous and radioactive species, (3) dewatering, (4) drying, and (5) solidification/stabilization. The first two options were eliminated from detailed consideration because they did not meet the system requirements. A quantitative evaluation clearly showed that, based on system constraints and project objectives, either dewatering or drying the mixed waste sludge was superior to the solidification/stabilization process option. The ultimate choice between the drying and the dewatering options will be made on the basis of a technical evaluation of the relative merits of proposals submitted by potential subcontractors

  11. Glassy slag: A complementary waste form to homogeneous glass for the implementation of MAWS in treating DOE low level/mixed wastes

    International Nuclear Information System (INIS)

    Feng, X.; Ordaz, G.; Krumrine, P.

    1994-01-01

    Glassy slag waste forms are being developed to complement glass waste forms in implementing the Minimum Additive Waste Stabilization (MAWS) Program for supporting DOE's environmental restoration efforts. These glassy slags are composed of various metal oxide crystalline phases embedded in an alumino-silicate glass phase. The slags are appropriate final waste forms for waste streams that contain large amounts of scrap metals and elements with low solubilities in glass, and that have low-flux contents. Homogeneous glass waste forms are appropriate for wastes with sufficient fluxes and low metal contents. Therefore, utilization of both glass and glassy slag waste forms will make vitrification technology applicable to the treatment of a much larger range of radioactive and mixed wastes. The MAWS approach was a plied to glassy slags by blending multiple waste streams to produce the final waste form, minimizing overall waste form volume and reducing costs. The crystalline oxide phases formed in the glassy slags can be specially formulated so that they are very durable and contain hazardous and radioactive elements in their lattice structures. The Structural Bond Strength (SBS) Model was used to predict the chemical durability of the product from the slag composition so that optimized slag compositions could be obtain with a limited number of crucible melts and testing

  12. Incineration/vitrification of radioactive wastes and combustion of pyrolysis gases in thermal plasmas

    International Nuclear Information System (INIS)

    Girold, Ch.

    1997-03-01

    Two thermal plasma processes used for incineration of radioactive technological wastes (cellulose, plastics, rubber...) have been investigated. First, the different types of radioactive wastes are presented, with a special attention to those which may benefit from a high temperature thermal treatment. The most significant thermal plasma processes, suitable for this goal, are described. Then, the author deals with the post-combustion, in an oxygen plasma jet reactor, of gases from burnable radioactive waste pyrolysis. An experimental planning method as been used to evaluate the combustion performances in the reactor, with a wide range of gas composition and running parameters such as oxygen excess and electrical power. The results of a modeling of kinetics, based on 116 chemicals reactions between 25 species, are compared with experimental values. Finally, an oxygen plasma reactor where the arc is transferred on a basalt melt is experimented. The efficiency of the combustion and the homogeneity of the glass are discussed. The volatility of some glass elements and tracers added to the wastes is also approached in two different ways: by post-trial material balance and by an optical emission spectroscopic method. The author built a diagnostic method that allows the following versus time of the metallic vapours above the melt. (author)

  13. Mixed waste paper as a fuel

    International Nuclear Information System (INIS)

    Kersletter, J.D.; Lyons, J.K.

    1991-01-01

    A successful recycling program requires several components: education and promotion, convenient collection service, and most importantly, a market for collected materials. In Washington state, domestic markets currently have, or are building, the capacity to use most of the glass, newsprint, aluminum, tin cans, and corrugated materials that are collected. Unfortunately, markets for mixed waste paper (MWP), a major component of the state's solid waste stream, have been slow to develop and are unable to absorb the tremendous volumes of material generated. The American Paper Stock Institute classifies MWP as low grade paper such as magazines, books, scrap paper, non-corrugated cardboard (boxboard/chipboard), and construction paper. When viewed as part of a curbside collection program MWP consists primarily of catalogs, binder paper, magazines, brochures, junk mail, cereal boxes, and other household packaging items. A comprehensive analysis of Washington State's solid waste stream showed that during 1988, Washington citizens generated approximately 460,000 tons of mixed waste paper. No small amount, this is equivalent to more than 10% of the total solid waste generated in the state, and is expected to increase. Current projections of MWP generation rates indicated that Washington citizens could discard as much as 960,000 tons of MWP by the year 2010 making it one of the single largest components of the state's solid waste stream. This paper reports on the use of MWP as fuel source

  14. Readiness review plan for the in situ vitrification demonstration of Seepage Pit 1 in Waste Area Grouping 7

    International Nuclear Information System (INIS)

    1995-05-01

    A treatability study is planned that encompasses the application of in situ vitrification (ISV) to at least two segments of the Oak Ridge National Laboratory Seepage Pit I during the third quarter of fiscal year 1995. Before the treatability study can be initiated, the proposed activity must be subjected to an Operational Readiness Review (ORR). ORR is a structured methodology of determining readiness to proceed as outlined in Martin Marietta Energy Systems, Inc. (Energy Systems), Environmental Restoration Waste Management Procedure ER/C-P1610, which provides Energy Systems organizations assurance that the work to be performed is consistent with management's expectations and that the subject activity is ready to proceed safely. The readiness review plan provides details of the review plan overview and the scope of work to be performed. The plan also identifies individuals and position responsibilities for implementing the activity. The management appointed Readiness Review Board (RRB) has been identified. A Field Readiness Review Team (FRT), a management appointed multidisciplinary group, has been established (1) to evaluate the ISV treatability study, (2) to identify and assemble supporting objective evidences of the readiness to proceed, and (3) to assist the team leader in presenting the evidences to the RRB. A major component of RRB is the formulation of readiness review criteria months before the operation. A comprehensive readiness review tree (a positive logic tree) is included, which identifies the activities required for the development of the readiness criteria. The readiness review tree serves as a tool to prevent the omission of an item that could affect system performance. All deficiencies identified in the review will be determined as prestart findings and must be resolved before the project is permitted to proceed. The final approval of the readiness to proceed will be the decision of RRB

  15. Bioprocessing scenarios for mixed hazardous waste

    International Nuclear Information System (INIS)

    Wolfram, J.H.; Rogers, R.D.

    1994-01-01

    The potential of biological processing of mixed hazardous waste has not been determined. However, the use of selected microorganisms for the degradation and/or detoxification of hazardous organic compounds is gaining wide acceptance as an alternative waste treatment technology. The isolation of a unique strain of Pseudomonas Putida Idaho seems well adapted to withstand the demands of the input stream comprised of liquid scintillation waste. This paper describes the results from the continuous processing of a mixture comprised of p-xylene and surfactant as well as commercial liquid scintillation formulations. The two formulations tested contained xylene and pseudocumene as the solvent base. The process is now at the demonstration phase at one of DOE's facilities which has a substantial amount of stored waste of this type. The system at the DOE facility is comprised of two CSTR units in series

  16. Vitrification of lead-rich solid ashes from incineration of hazardous industrial wastes

    International Nuclear Information System (INIS)

    Kavouras, P.; Kaimakamis, G.; Ioannidis, Th. A.; Kehagias, Th.; Komninou, Ph.; Kokkou, S.; Pavlidou, E.; Antonopoulos, I.; Sofoniou, M.; Zouboulis, A.; Hadjiantoniou, C.P.; Nouet, G.; Prakouras, A.; Karakostas, Th.

    2003-01-01

    Lead-rich solid industrial wastes were vitrified by the addition of glass formers in various concentrations, to produce non-toxic vitreous stabilized products that can be freely disposed or used as construction materials. Toxicity of both the as-received industrial solid waste and the stabilized products was determined using standard leaching test procedures. The chemically stable vitreous products were subjected to thermal annealing in order to investigate the extent of crystal separation that could occur during cooling of large pieces of glass. Leaching tests were repeated to investigate the relation between annealing process and chemical stability. X-ray, scanning and transmission electron microscopy techniques were employed to identify the microstructure of stabilized products before and after thermal treatment. Relation between synthesis and processing, chemical stability and microstructure was investigated

  17. Nuclear wastes: vitrification and long term behaviour. Information available on Internet sites

    International Nuclear Information System (INIS)

    Macias, R.M.

    2004-01-01

    This document reports an information search on the long-term behaviour of vitrified nuclear wastes. This search has been performed on the Internet and this report presents 21 forms corresponding to the 21 relevant Internet web sites. These forms contain the name of the organization, its country, its Internet address, its language, a brief description of the available information, the concerned public (general public, scholars and students, scientific community)

  18. Materials performance in a high-level radioactive waste vitrification system

    International Nuclear Information System (INIS)

    Imrich, K.J.; Chandler, G.T.

    1996-01-01

    The Defense Waste Processing Facility (DWPF) is a Department of Energy Facility designed to vitrify highly radioactive waste. An extensive materials evaluation program has been completed on key components in the DWPF after twelve months of operation using nonradioactive simulated wastes. Results of the visual inspections of the feed preparation system indicate that the system components, which were fabricated from Hastelloy C-276, should achieve their design lives. Significant erosion was observed on agitator blades that process glass frit slurries; however, design modifications should mitigate the erosion. Visual inspections of the DWPF melter top head and off gas components, which were fabricated from Inconel 690, indicated that varying degrees of degradation occurred. Most of the components will perform satisfactorily for their two year design life. The components that suffered significant attack were the borescopes, primary film cooler brush, and feed tubes. Changes in the operation of the film cooler brush and design modifications to the feed tubes and borescopes is expected to extend their service lives to two years. A program to investigate new high temperature engineered materials and alloys with improved oxidation and high temperature corrosion resistance will be initiated

  19. Vitrification of galvanic solid wastes: solutions for the east area of Sao Paulo, Brazil

    International Nuclear Information System (INIS)

    Mattos, Cleiton dos Santos; Castanho, Sonia Regina Homem de Mello

    2011-01-01

    Galvanic solid waste have elevated levels of heavy metals and usually are stocked in the industry, creating a worrisome environmental liabilities. This disturbing fact is aggravated in areas densely populated as the area east of Sao Paulo, which has a pole of industrial electroplating of chrome. The present paper, we describe and provide a technological option for the disposal of waste generated by this activity using techniques that allow the incorporation of these in a glass matrix. The wastes were characterized by XRF, EDS, ICP-AES, AAS, DTA/TGA, XRD and SEM-FEG and embedded in glass and frits made from the system SiO_-CaO-Na_O, with additions of up to 30% by weight. The results of the analysis of residues showed the majority presence of Ni, Cr, B, Cu, Ca and S. The resulting glasses showed that heavy metals were incorporated into its structure and probably replacing the Ca and Na. In addition, the products showed specific colors indicating the possibility of use in some segments of manufacturing in ceramics with glazes, loading and pigments. (author)

  20. Remote waste handling and feed preparation for Mixed Waste Management

    International Nuclear Information System (INIS)

    Couture, S.A.; Merrill, R.D.; Densley, P.J.

    1995-05-01

    The Mixed Waste Management Facility (MWMF) at the Lawrence Livermore National Laboratory (LLNL) will serve as a national testbed to demonstrate mature mixed waste handling and treatment technologies in a complete front-end to back-end --facility (1). Remote operations, modular processing units and telerobotics for initial waste characterization, sorting and feed preparation have been demonstrated at the bench scale and have been selected for demonstration in MWMF. The goal of the Feed Preparation design team was to design and deploy a robust system that meets the initial waste preparation flexibility and productivity needs while providing a smooth upgrade path to incorporate technology advances as they occur. The selection of telerobotics for remote handling in MWMF was made based on a number of factors -- personnel protection, waste generation, maturity, cost, flexibility and extendibility. Modular processing units were selected to enable processing flexibility and facilitate reconfiguration as new treatment processes or waste streams are brought on line for demonstration. Modularity will be achieved through standard interfaces for mechanical attachment as well as process utilities, feeds and effluents. This will facilitate reconfiguration of contaminated systems without drilling, cutting or welding of contaminated materials and with a minimum of operator contact. Modular interfaces also provide a standard connection and disconnection method that can be engineered to allow convenient remote operation

  1. Laboratory-scale vitrification and leaching of Hanford high-level waste for the purpose of simulant and glass property models validation

    International Nuclear Information System (INIS)

    Morrey, E.V.; Elliott, M.L.; Tingey, J.M.

    1993-02-01

    The Hanford Waste Vitrification Plant (HWVP) is being built to process the high-level and TRU waste into canistered glass logs for disposal in a national repository. Testing programs have been established within the Project to verify process technology using simulated waste. A parallel testing program with actual radioactive waste is being performed to confirm the validity of using simulates and glass property models for waste form qualification and process testing. The first feed type to be processed by and the first to be tested on a laboratory-scale is pretreated neutralized current acid waste (NCAW). The NCAW is a neutralized high-level waste stream generated from the reprocessing of irradiated nuclear fuel in the Plutonium and Uranium Extraction (PUREX) Plant at Hanford. As part of the fuel reprocessing, the high-level waste generated in PUREX was denitrated with sugar to form current acid waste (CAW). Sodium hydroxide and sodium nitrite were added to the CAW to minimize corrosion in the tanks, thus yielding neutralized CAW. The NCAW contains small amounts of plutonium, fission products from the irradiated fuel, stainless steel corrosion products, and iron and sulfate from the ferrous sulfamate reductant used in the PUREX process. This paper will discuss the results and status of the laboratory-scale radioactive testing

  2. Mixed waste management in Washington and the Northwest Compact Region

    International Nuclear Information System (INIS)

    Carlin, E.M.

    1988-01-01

    The state of Washington's concerns about the management of mixed waste have evolved over the past year. One concern that receives increasing attention is the Northwest Compact Region's need to plan for disposal of its own mixed waste. An informal survey of the region's potential mixed waste generators has indicated that mixed waste volumes are low. However, the opening of a disposal facility may result in increased waste volumes. A preliminary proposal for such a facility has been reviewed by the federal and state agencies that dually regulate mixed waste. Initial conclusions reached by the regulators are presented

  3. Mixed waste chemical compatibility with packaging components

    International Nuclear Information System (INIS)

    Nigrey, P.J.; Conroy, M.; Blalock, L.B.

    1994-01-01

    In this paper, a chemical compatibility testing program for packaging of mixed wastes at will be described. We will discuss the choice of four y-radiation doses, four time durations, four temperatures and four waste solutions to simulate the hazardous waste components of mixed wastes for testing materials compatibility of polymers. The selected simulant wastes are (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; and (4) a mixture of ketones. A selection of 10 polymers with anticipated high resistance to one or more of these types of environments are proposed for testing as potential liner or seal materials. These polymers are butadiene acrylonitrile copolymer, cross-linked polyethylene, epichlorhyarin, ethylene-propylene rubber, fluorocarbon, glass-filled tetrafluoroethylene, high-density poly-ethylene, isobutylene-isoprene copolymer, polypropylene, and styrene-butadiene rubber. We will describe the elements of the testing plan along with a metric for establishing time resistance of the packaging materials to radiation and chemicals

  4. Vitrification processes for fission product solutions

    International Nuclear Information System (INIS)

    Bonniaud, R.; Jouan, A.; Moncouyoux, J.P.; Sombret, C.

    1982-10-01

    The different processes for fission product vitrification in the world are reviewed. Continuous or discontinuous processes, induction or arc heating, in can melting or casting, tests with radioactive or simulated wastes and industrial realizations are described [fr

  5. Engineering-scale in situ vitrification tests of simulated Oak Ridge National Laboratory buried wastes

    International Nuclear Information System (INIS)

    1996-12-01

    As part of the Comprehensive Environmental Response, Compensation, and Liability Act process for remediation of Waste Area Grouping (WAG) 6 at Oak Ridge National Laboratory, a public meeting was held on the proposed plan. It was recognized that contaminant releases from WAG 6 posed minimal potential risk to the public and environment. The US Department of Energy (DOE) in conjunction with the US Environmental Protection Agency and the Tennessee Department of Environment and Conservation agreed to defer remedial action at WAG 6 until higher risk release sites were first remediated

  6. Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report

    International Nuclear Information System (INIS)

    Herborn, D.I.

    1991-10-01

    The requirements for Westinghouse Hanford independent review of the Preliminary Safety Analysis Report (PSAR) are contained in Section 1.0, Subsection 4.3 of WCH-CM-4-46. Specifically, this manual requires the following: (1) Formal functional reviews of the HWVP PSAR by the future operating organization (HWVP Operations), and the independent review organizations (HWVP and Environmental Safety Assurance, Environmental Assurance, and Quality Assurance); and (2) Review and approval of the HWVP PSAR by the Tank Waste Disposal (TWD) Subcouncil of the Safety and Environmental Advisory Council (SEAC), which provides independent advice to the Westinghouse Hanford President and executives on matters of safety and environmental protection. 7 refs

  7. Volatilization behavior of semivolatile elements in vitrification of high-level liquid waste

    International Nuclear Information System (INIS)

    Igarashi, Hiroshi; Kato, Koh; Takahashi, Takeshi

    1991-11-01

    The effect of temperature on the volatilization of ruthenium, technetium, and selenium was observed in calcination experiments with simulated high-level liquid waste. Technetium and selenium were more volatile as calcining temperature increased. Ruthenium was less volatile when temperature exceeded 300degC. More than 80% of ruthenium that volatilized from room temperature to 500degC occurred between 200 and 300degC. A small amount of ruthenium volatilized above 300degC as well as below 135degC. (author)

  8. Design of a mixing system for simulated high-level nuclear waste melter feed slurries

    International Nuclear Information System (INIS)

    Peterson, M.E.; McCarthy, D.; Muhlstein, K.D.

    1986-03-01

    The Nuclear Waste Treatment Program development program consists of coordinated nonradioactive and radioactive testing combined with numerical modeling of the process to provide a complete basis for design and operation of a vitrification facility. The radioactive demonstration tests of equipment and processes are conducted before incorporation in radioactive pilot-scale melter systems for final demonstration. The mixing system evaluation described in this report was conducted as part of the nonradioactive testing. The format of this report follows the sequence in which the design of a large-scale mixing system is determined. The initial program activity was concerned with gaining an understanding of the theoretical foundation of non-Newtonian mixing systems. Section 3 of this report describes the classical rheological models that are used to describe non-Newtonian mixing systems. Since the results obtained here are only valid for the slurries utilized, Section 4, Preparation of Simulated Hanford and West Valley Slurries, describes how the slurries were prepared. The laboratory-scale viscometric and physical property information is summarized in Section 5, Laboratory Rheological Evaluations. The bench-scale mixing evaluations conducted to define the effects of the independent variables described above on the degree of mixing achieved with each slurry are described in Section 6. Bench-scale results are scaled-up to establish engineering design requirements for the full-scale mixing system in Section 7. 24 refs., 37 figs., 44 tabs

  9. Development of glass compositions with 9% waste content for the vitrification of high-level waste from LWR nuclear reactors

    International Nuclear Information System (INIS)

    Lakatos, T.

    1979-10-01

    Reduction of the contents of waste in glass from 20-25% to 9% causes a decrease of the leaching resistance of the glass. The addition of Zn0 reduces the leaching values by a factor of approximately 10. The crystallized glass ceramics have a lower coefficient of thermal expansion than glassy waste bodies. The separation of the phase which contains Mo occurs during heat treatment. The amount of separated Mo is lower for low alkali sac type (Si0 2 - A1 2 0 3 -Ca0 system) of glasses by a factor of approximately 50. All the glasses were prepared with simulated waste composition. (GBn.)

  10. Study on fused/cast AZS refractories for deployment in vitrification of radioactive waste effluents

    Science.gov (United States)

    Sengupta, Pranesh; Mishra, R. K.; Soudamini, N.; Sen, D.; Mazumder, S.; Kaushik, C. P.; Ajithkumar, T. G.; Banerjee, K.

    2015-12-01

    'Fused/cast Al2O3-ZrO2-SiO2 (FC-AZS)' is being considered as 'glass contact refractory' within ceramic melters, to be used for nuclear waste immobilization. Microstructural analyses reveal random distributions of baddeleyite (ZrO2) within aluminosilicate (Al2SiO5) matrix. 27Al and 29Si NMR data suggest that within aluminosilicate matrix Al occurs in both 4- and 6-fold co-ordinations whereas Si prefers a 4-fold environment. Polydispersity of pores has been studied with small-angle neutron scattering (SANS) technique. Corrosion rates of FC-AZS within 6 M HNO3, simulated wastes (500 h exposure), and borosilicate melt (975 °C, 800 h exposure) are found to be 0.38 × 103 μmy-1, 0.13 × 103 μmy-1 and 4.75 × 103 μmy-1 respectively. A comparison of chemical interaction data clearly suggests that FC-AZS exhibits better chemical durability than AZC refractory (Al2O3-ZrO2-Cr2O3, also used for similar purpose). Thermal cycling studies indicate that FC-AZS retains structural integrity (including compressive strength and density) even up to 20 cycles.

  11. Measurement of cesium emissions during the vitrification of simulated high level radioactive waste

    International Nuclear Information System (INIS)

    Zamecnik, J.R.; Miller, D.H.; Carter, J.T.

    1992-01-01

    In the Defense Waste Processing Facility at the Savannah River Site, it is desired to eliminate a startup test that would involve adding small amounts of radioactive cesium-137 to simulated high-level waste. In order to eliminate this test, a reliable method for measuring non-radioactive cesium in the offgas system from the glass melter is required. From a pilot scale melter system, offgas particulate samples were taken on filter paper media and analyzed by Inductively Coupled Plasma-Mass Spectrometry (ICP-MS). The ICPMS method proved to be sufficiently sensitive to measure cesium quantities as low as 0.135 μg, with the sensitivity being limited by the background cesium present in the filter paper. Typical particulate loadings ranged from 800 μg of cesium. This sensitivity allowed determination of cesium decontamination factors for four of the five major components of the offgas system. The decontamination factors measured experimentally compared favorably with the process design basis values

  12. Recommendations for continuous emissions monitoring of mixed waste incinerators

    International Nuclear Information System (INIS)

    Quigley, G.P.

    1992-01-01

    Considerable quantities of incinerable mixed waste are being stored in and generated by the DOE complex. Mixed waste is defined as containing a hazardous component and a radioactive component. At the present time, there is only one incinerator in the complex which has the proper TSCA and RCRA permits to handle mixed waste. This report describes monitoring techniques needed for the incinerator

  13. Hanford Central Waste Complex: Radioactive mixed waste storage facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-10-01

    The Hanford Site is owned by the US Government and operated by the US Department of Energy Field Office, Richland. The Hanford Site manages and produces dangerous waste and mixed waste (containing both radioactive and dangerous components). The dangerous waste is regulated in accordance with the Resource Conversation and Recovery Act of 1976 and the State of Washington Hazardous Waste Management Act of 1976. The radioactive component of mixed waste is interpreted by the US Department of Energy to be regulated under the Atomic Energy Act of 1954; the nonradioactive dangerous component of mixed waste is interpreted to be regulated under the Resource Conservation and Recovery Act of 1976 and Washington Administrative Code 173--303. Westinghouse Hanford Company is a major contractor to the US Department of Energy Field Office, Richland and serves as co-operator of the Hanford Central Waste Complex. The Hanford Central Waste Complex is an existing and planned series of treatment, storage, and/or disposal units that will centralize the management of solid waste operations at a single location on the Hanford facility. The Hanford Central Waste Complex units include the Radioactive Mixed Waste Storage Facility, the unit addressed by this permit application, and the Waste Receiving and Processing Facility. The Waste Receiving and Processing Facility is covered in a separate permit application submittal

  14. Evaluation of chemical stability of vitrification media for radioactive waste products

    International Nuclear Information System (INIS)

    Barkatt, A.; Simmons, J.H.; Macedo, P.B.

    1981-01-01

    Test methods and test results concerning the measurement of chemical durability of glass media proposed for nuclear waste fixation are described. In order to develop predictive models and risk calculations, the release rates of individual components are measured. The results are used to determine matrix dissolution rates, possible transport of components through the matrix, and chemical and physical corrosion mechanisms. Measurements on model borosilicate and high silica glass fixation media are reported and discussed in terms of layer formation, approach to steady state, interaction of polyvalent ions with the dealkalised layer, structural disintegration of the layer, and the effects of glass composition and of environmental conditions (temperature, leachant composition and pH, γ dose). The extrapolation of short term laboratory tests to long time storage conditions and the use of such extrapolation in predicting safe upper limits for the release rates of components of the glass are described. (author)

  15. Oak Ridge National Laboratory West End Treatment Facility simulated sludge vitrification demonstration, Revision 1

    International Nuclear Information System (INIS)

    Cicero, C.A.; Bickford, D.F.; Bennert, D.M.; Overcamp, T.J.

    1994-01-01

    Technologies are being developed by the US Department of Energy's (DOE) Nuclear Facility sites to convert hazardous and mixed wastes to a form suitable for permanent disposal. Vitrification, which has been declared the Best Demonstrated Available Technology for high-level radioactive waste disposal by the EPA, is capable of producing a highly durable wasteform that minimizes disposal volumes through organic destruction, moisture evaporation, and porosity reduction. However, this technology must be demonstrated over a range of waste characteristics, including compositions, chemistries, moistures, and physical characteristics to ensure that it is suitable for hazardous and mixed waste treatment. These wastes are typically wastewater treatment sludges that are categorized as listed wastes due to the process origin or organic solvent content, and usually contain only small amounts of hazardous constituents. The Oak Ridge National Laboratory's (ORNL) West End Treatment Facility's (WETF) sludge is considered on of these representative wastes. The WETF is a liquid waste processing plant that generates sludge from the biodenitrification and precipitation processes. An alternative wasteform is needed since the waste is currently stored in epoxy coated carbon steel tanks, which have a limited life. Since this waste has characteristics that make it suitable for vitrification with a high likelihood of success, it was identified as a suitable candidate by the Mixed Waste Integrated Program (MWIP) for testing at CU. The areas of special interest with this sludge are (1) minimum nitrates, (2) organic destruction, and (3) waste water treatment sludges containing little or no filter aid

  16. Strategy for managing mixed waste at a plant site

    International Nuclear Information System (INIS)

    Fentiman, A.

    1991-01-01

    No waste disposal site is currently accepting mixed waste, but facilities across the country continue to generate it. The waste manager at each site is faced with two problems: how to manage the mixed waste already on-site and how to minimize the amount of new waste generated. A strategy has been developed to address each problem. A key element of the strategy is a building-by-building survey of the site. The survey provides information on how and where mixed waste is generated and stored. This paper describes a method for planning and conducting a site-wide mixed-waste survey. It then outlines approaches to managing existing mixed waste and to minimizing mixed-waste generation using information from the survey

  17. One System Integrated Project Team: Retrieval And Delivery Of The Hanford Tank Wastes For Vitrification In The Waste Treatment Plant

    International Nuclear Information System (INIS)

    Harp, Benton J.; Kacich, Richard M.; Skwarek, Raymond J.

    2012-01-01

    The One System Integrated Project Team (IPT) was formed in late 2011 as a way for improving the efficiency of delivery and treatment of highly radioactive waste stored in underground tanks at the U.S. Department of Energy's (DOE's) 586-square-mile Hanford Site in southeastern Washington State. The purpose of the One System IPT is to improve coordination and integration between the Hanford's Waste Treatment Plant (WTP) contractor and the Tank Operations Contractor (TOC). The vision statement is: One System is a WTP and TOC safety conscious team that, through integrated management and implementation of risk-informed decision and mission-based solutions, will enable the earliest start of safe and efficient treatment of Hanford's tank waste, to protect the Columbia River, environment and public. The IPT is a formal collaboration between Bechtel National, Inc. (BNI), which manages design and construction of the WTP for the U.S. Department of Energy's Office of River Protection (DOEORP), and Washington River Protection Solutions (WRPS), which manages the TOC for ORP. More than fifty-six (56) million gallons of highly radioactive liquid waste are stored in one hundred seventy-seven (177) aging, underground tanks. Most of Hanford's waste tanks - one hundred forty-nine (149) of them - are of an old single-shell tank (SST) design built between 1944 and 1964. More than sixty (60) of these tanks have leaked in the past, releasing an estimated one million gallons of waste into the soil and threatening the nearby Columbia River. There are another twenty-eight (28) new double-shelled tanks (DSTs), built from 1968 to 1986, that provide greater protection to the environment. In 1989, DOE, the U.S. Environmental Protection Agency (EPA), and the Washington State Department of Ecology (Ecology) signed a landmark agreement that required Hanford to comply with federal and state environmental standards. It also paved the way for agreements that set deadlines for retrieving the tank

  18. One System Integrated Project Team: Retrieval And Delivery Of The Hanford Tank Wastes For Vitrification In The Waste Treatment Plant

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Benton J. [Department of Energy, Office of River Protection, Richland, Washington (United States); Kacich, Richard M. [Bechtel National, Inc., Richland, WA (United States); Skwarek, Raymond J. [Washington River Protection Solutions LLC, Richland, WA (United States)

    2012-12-20

    The One System Integrated Project Team (IPT) was formed in late 2011 as a way for improving the efficiency of delivery and treatment of highly radioactive waste stored in underground tanks at the U.S. Department of Energy's (DOE's) 586-square-mile Hanford Site in southeastern Washington State. The purpose of the One System IPT is to improve coordination and integration between the Hanford's Waste Treatment Plant (WTP) contractor and the Tank Operations Contractor (TOC). The vision statement is: One System is a WTP and TOC safety conscious team that, through integrated management and implementation of risk-informed decision and mission-based solutions, will enable the earliest start of safe and efficient treatment of Hanford's tank waste, to protect the Columbia River, environment and public. The IPT is a formal collaboration between Bechtel National, Inc. (BNI), which manages design and construction of the WTP for the U.S. Department of Energy's Office of River Protection (DOEORP), and Washington River Protection Solutions (WRPS), which manages the TOC for ORP. More than fifty-six (56) million gallons of highly radioactive liquid waste are stored in one hundred seventy-seven (177) aging, underground tanks. Most of Hanford's waste tanks - one hundred forty-nine (149) of them - are of an old single-shell tank (SST) design built between 1944 and 1964. More than sixty (60) of these tanks have leaked in the past, releasing an estimated one million gallons of waste into the soil and threatening the nearby Columbia River. There are another twenty-eight (28) new double-shelled tanks (DSTs), built from 1968 to 1986, that provide greater protection to the environment. In 1989, DOE, the U.S. Environmental Protection Agency (EPA), and the Washington State Department of Ecology (Ecology) signed a landmark agreement that required Hanford to comply with federal and state environmental standards. It also paved the way for agreements that set deadlines for retrieving the tank

  19. One System Integrated Project Team: Retrieval and Delivery of Hanford Tank Wastes for Vitrification in the Waste Treatment Plant - 13234

    International Nuclear Information System (INIS)

    Harp, Benton J.; Kacich, Richard M.; Skwarek, Raymond J.

    2013-01-01

    The One System Integrated Project Team (IPT) was formed in late 2011 as a way for improving the efficiency of delivery and treatment of highly radioactive waste stored in underground tanks at the U.S. Department of Energy's (DOE's) 586-square-mile Hanford Site in southeastern Washington State. The purpose of the One System IPT is to improve coordination and integration between the Hanford's Waste Treatment Plant (WTP) contractor and the Tank Operations Contractor (TOC). The vision statement is: One System is a WTP and TOC safety-conscious team that, through integrated management and implementation of risk-informed decision and mission-based solutions, will enable the earliest start of safe and efficient treatment of Hanford's tank waste, to protect the Columbia River, environment and public. The IPT is a formal collaboration between Bechtel National, Inc. (BNI), which manages design and construction of the WTP for the U.S. Department of Energy's Office of River Protection (DOEORP), and Washington River Protection Solutions (WRPS), which manages the TOC for ORP. More than fifty-six (56) million gallons of highly radioactive liquid waste are stored in one hundred seventy-seven (177) aging, underground tanks. Most of Hanford's waste tanks - one hundred forty-nine (149) of them - are of an old single-shell tank (SST) design built between 1944 and 1964. More than sixty (60) of these tanks have leaked in the past, releasing an estimated one million gallons of waste into the soil and threatening the nearby Columbia River. There are another twenty-eight (28) new double-shelled tanks (DSTs), built from 1968 to 1986, that provide greater protection to the environment. In 1989, DOE, the U.S. Environmental Protection Agency (EPA), and the Washington State Department of Ecology (Ecology) signed a landmark agreement that required Hanford to comply with federal and state environmental standards. It also paved the way for agreements that set deadlines for retrieving the tank

  20. One System Integrated Project Team: Retrieval and Delivery of Hanford Tank Wastes for Vitrification in the Waste Treatment Plant - 13234

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Benton J. [U.S. Department of Energy, Office of River Protection, Post Office Box 550, Richland, Washington 99352 (United States); Kacich, Richard M. [Bechtel National, Inc., 2435 Stevens Center Place, Richland, Washington 99354 (United States); Skwarek, Raymond J. [Washington River Protection Solutions LLC, Post Office Box 850, Richland, Washington 99352 (United States)

    2013-07-01

    The One System Integrated Project Team (IPT) was formed in late 2011 as a way for improving the efficiency of delivery and treatment of highly radioactive waste stored in underground tanks at the U.S. Department of Energy's (DOE's) 586-square-mile Hanford Site in southeastern Washington State. The purpose of the One System IPT is to improve coordination and integration between the Hanford's Waste Treatment Plant (WTP) contractor and the Tank Operations Contractor (TOC). The vision statement is: One System is a WTP and TOC safety-conscious team that, through integrated management and implementation of risk-informed decision and mission-based solutions, will enable the earliest start of safe and efficient treatment of Hanford's tank waste, to protect the Columbia River, environment and public. The IPT is a formal collaboration between Bechtel National, Inc. (BNI), which manages design and construction of the WTP for the U.S. Department of Energy's Office of River Protection (DOEORP), and Washington River Protection Solutions (WRPS), which manages the TOC for ORP. More than fifty-six (56) million gallons of highly radioactive liquid waste are stored in one hundred seventy-seven (177) aging, underground tanks. Most of Hanford's waste tanks - one hundred forty-nine (149) of them - are of an old single-shell tank (SST) design built between 1944 and 1964. More than sixty (60) of these tanks have leaked in the past, releasing an estimated one million gallons of waste into the soil and threatening the nearby Columbia River. There are another twenty-eight (28) new double-shelled tanks (DSTs), built from 1968 to 1986, that provide greater protection to the environment. In 1989, DOE, the U.S. Environmental Protection Agency (EPA), and the Washington State Department of Ecology (Ecology) signed a landmark agreement that required Hanford to comply with federal and state environmental standards. It also paved the way for agreements that set deadlines

  1. MWIP: Surrogate formulations for thermal treatment of low-level mixed waste

    International Nuclear Information System (INIS)

    Bostick, W.D.; Hoffmann, D.P.; Stevenson, R.J.; Richmond, A.A.; Bickford, D.F.

    1994-01-01

    The category of sludges, filter cakes, and other waste processing residuals represent the largest volume of low-level mixed (hazardous and radioactive) wastes within the US Department of Energy (DOE) complex. Treatment of these wastes to minimize the mobility of contaminants, and to eliminate the presence of free water, is required under the Federal Facility Compliance Act agreements between DOE and the Environmental Protection Agency. In the text, we summarize the currently available data for several of the high priority mixed-waste sludge inventories within DOE. Los Alamos National Laboratory TA-50 Sludge and Rocky Flats Plant By-Pass Sludge are transuranic (TRU)-contaminated sludges that were isolated with the use of silica-based filter aids. The Oak Ridge Y-12 Plant West End Treatment Facility Sludge is predominantly calcium carbonate and biomass. The Oak Ridge K-25 Site Pond Waste is a large-volume waste stream, containing clay, silt, and other debris in addition to precipitated metal hydroxides. We formulate ''simulants'' for the waste streams described above, using cerium oxide as a surrogate for the uranium or plutonium present in the authentic material. Use of nonradiological surrogates greatly simplifies material handling requirements for initial treatability studies. The use of synthetic mixtures for initial treatability testing will facilitate compositional variation for use in conjunction with statistical design experiments; this approach may help to identify any ''operating window'' limitations. The initial treatability testing demonstrations utilizing these ''simulants'' will be based upon vitrification, although the materials are also amenable to testing grout-based and other stabilization procedures. After the feasibility of treatment and the initial evaluation of treatment performance has been demonstrated, performance must be verified using authentic samples of the candidate waste stream

  2. Development of vitrification line technology and the manufacture of equipment

    International Nuclear Information System (INIS)

    Alexa, J.

    1989-01-01

    The development is described of technology and the production of equipment for the vitrification of liquid radioactive wastes. For vitrification, frit Frita F270 is used containing up to 20% titanium and featuring a corrosion effect lower by one order than that of lead glass. The liquid waste is discharged in a measuring tank where it is mixed with formic acid. It is then pumped into an evaporator. Breed vapor is carried via a condenser to a condensate tank. The evaporator concentrate is transported to a homogenizer where it is gradually mixed with Frita. The viscous mush thus produced is carried into a furnace where the remaining water is evaporated. The furnace decontamination factor is 10 2 to 10 3 . At a temperature of up to 1,050 degC the frit melts and is discharged into a case. Currently, technology has been developed of mush preparation and the design has been completed of a vitrification furnace featuring remote lid opening and closing, and of equipment for processing furnace emissions. (J.B.). 3 figs., 1 tab., 1 ref

  3. Performance evaluation of rotating pump jet mixing of radioactive wastes in Hanford Tanks 241-AP-102 and -104

    International Nuclear Information System (INIS)

    Onishi, Y.; Recknagle, K.P.

    1998-07-01

    The purpose of this study was to confirm the adequacy of a single mixer pump to fully mix the wastes that will be stored in Tanks 241-AP-102 and -104. These Hanford double-shell tanks (DSTs) will be used as staging tanks to receive low-activity wastes from other Hanford storage tanks and, in turn, will supply the wastes to private waste vitrification facilities for eventual solidification. The TEMPEST computer code was applied to Tanks AP-102 and -104 to simulate waste mixing generated by the 60-ft/s rotating jets and to determine the effectiveness of the single rotating pump to mix the waste. TEMPEST simulates flow and mass/heat transport and chemical reactions (equilibrium and kinetic reactions) coupled together. Section 2 describes the pump jet mixing conditions the authors evaluated, the modeling cases, and their parameters. Section 3 reports model applications and assessment results. The summary and conclusions are presented in Section 4, and cited references are listed in Section 5

  4. Determining how much mixed waste will require disposal

    International Nuclear Information System (INIS)

    Kirner, N.P.

    1990-01-01

    Estimating needed mixed-waste disposal capacity to 1995 and beyond is an essential element in the safe management of low-level radioactive waste disposal capacity. Information on the types and quantities of mixed waste generated is needed by industry to allow development of treatment facilities and by states and others responsible for disposal and storage of this type of low-level radioactive waste. The design of a mixed waste disposal facility hinges on a detailed assessment of the types and quantities of mixed waste that will ultimately require land disposal. Although traditional liquid scintillation counting fluids using toluene and xylene are clearly recognized as mixed waste, characterization of other types of mixed waste has, however, been difficult. Liquid scintillation counting fluids comprise most of the mixed waste generated and this type of mixed waste is generally incinerated under the supplemental fuel provisions of the Resource Conservation and Recovery Act (RCRA) Because there are no Currently operating mixed waste land disposal facilities, it is impossible to make projections of waste requiring land disposal based on a continuation of current waste disposal practices. Evidence indicates the volume of mixed waste requiring land disposal is not large, since generators are apparently storing these wastes. Surveys conducted to date confirm that relatively small volumes of commercially generated mixed waste volume have relied heavily oil generators' knowledge of their wastes. Evidence exists that many generators are confused by the differences between the Atomic Energy Act and the Resource Conservation and Recovery Act (RCRA) on the issue of when a material becomes a waste. In spite of uncertainties, estimates of waste volumes requiring disposal can be made. This paper proposes an eight-step process for such estimates

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

    International Nuclear Information System (INIS)

    1995-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

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

  7. Melter Feed Reactions at T ≤ 700°C for Nuclear Waste Vitrification

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Kai [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hrma, Pavel R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rice, Jarrett A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-07-23

    Batch reactions and phase transitions in a nuclear waste feed heated at 5 K min-1 up to 600°C were investigated by optical microscopy, scanning electron microscopy with energy dispersive X-ray spectrometer, and X-ray diffraction. Quenched samples were leached in deionized water at room temperature and 80°C to extract soluble salts and early glass-forming melt, respectively. To determine the content and composition of leachable phases, the leachates were analyzed by the inductively-coupled plasma spectroscopy. By ~400°C, gibbsite and borax lost water and converted to amorphous and intermediate crystalline phases. Between 400°C and 600°C, the sodium borate early glass-forming melt reacted with amorphous aluminum oxide and calcium oxide to form intermediate products containing Al and Ca. At ~600°C, half Na and B converted to the early glass-forming melt, and quartz began to dissolve in the melt.

  8. Measurement of cesium and mercury emissions from the vitrification of simulated high level radioactive waste

    International Nuclear Information System (INIS)

    Zamecnik, J.R.

    1992-01-01

    In the Defense Waste Processing Facility at the Savannah River Site, it is desired to measure non-radioactive cesium in the offgas system from the glass melter. From a pilot scale melter system, offgas particulate samples were taken on filter paper media and analyzed by Inductively Coupled Plasma-Mass Spectrometry (ICP-MS). The ICP-MS method proved to be sufficiently sensitive to measure cesium quantities as low as 0.135 μg, with the sensitivity being limited by the background cesium present in the filter paper. This sensitivity allowed determination of cesium decontamination factors for four of the five major components of the offgas system. In addition, total particulate measurements were also made. Measurements of mercury decontamination factors were made on the same equipment; the results indicate that most of the mercury in the offgas system probably exists as elemental mercury and HgCl 2 , with some HgO and Hg 2 Cl 2 . The decontamination factors determined for cesium, total particulate, and mercury all compared favorably with the design values

  9. Wastes vitrification by plasma torch: study of a glass formulation compatible with a wide range of B wastes; Vitrification des dechets par torche a plasma: recherche d'une formulation de verre compatible avec un large eventail de dechets B

    Energy Technology Data Exchange (ETDEWEB)

    Poitou, S.; Richaud, D.; Fiquet, O.; Gramondi, P.; Massit, H. [CEA Cadarache, Dept. d' Entreposage et de Stockage des Dechets, 13 - Saint-Paul-lez-Durance (France)

    2001-07-01

    Within the context of radioactive waste management, CEA has equipped itself with a 'PLASMARC' device. The central element of this device is a plasma torch treatment furnace. It has been implemented and validated for the vitrification of low level radioactive wastes. Meanwhile, the plasma torch presents potentially interests for immobilizing under an inert form in vitreous matrices, B wastes which are generally divided and of complex chemical composition. The application of this process to this type of wastes has been studied here. The obtained results show that with the plasma torch it is possible to make glasses with a high amount of silicon and aluminium oxide and which are adapted to the treatment / packaging of the B wastes. (O.M.)

  10. Formulation Efforts for Direct Vitrification of INEEL Blend Calcine Waste Simulate: Fiscal Year 2000

    International Nuclear Information System (INIS)

    Crum, Jarrod V; Vienna, John D; Peeler, David K

    2001-01-01

    The TFA uses a systematic process for developing its annual program that draws from the tanks science and technology development needs expressed by the five DOE tank waste sites. TFA's annual program development process is iterative and involves the following steps: Collection of site needs; Needs analysis; Development of technical responses and initial prioritization; Refinement of the program for the next fiscal year; Formulation of the Corporate Review Budget (CRB); Preparation of Program Execution Guidance (PEG) for the next FY Revision of the Multiyear Program Plan (MYPP). This document describes the outcomes of the first phase of this process, from collection of site needs to the initial prioritization of technical activities. The TFA received site needs in October- December 2000. A total of 170 site needs were received, an increase of 30 over the previous year. The needs were analyzed and integrated, where appropriate. Sixty-six distinct technical responses were drafted and prioritized. In addition, seven strategic tasks were approved to compete for available funding in FY 2002 and FY 2003. Draft technical responses were prepared and provided to the TFA Site Representatives and the TFA User Steering Group (USG) for their review and comment. These responses were discussed at a March 15, 2001, meeting where the TFA Management Team established the priority listing in preparation for input to the DOE Office of Science and Technology (OST) budget process. At the time of publication of this document, the TFA continues to finalize technical responses as directed by the TFA Management Team and clarify the intended work scopes for FY 2002 and FY 2003

  11. Waste-Mixes Study for space disposal

    International Nuclear Information System (INIS)

    McCallum, R.F.; Blair, H.T.; McKee, R.W.; Silviera, D.J.; Swanson, J.L.

    1983-01-01

    The Wastes Mixes Study is a component of Cy-1981 and 1982 research activities to determine if space disposal could be a feasible complement to geologic disposal for certain high-level (HLW) and transuranic wastes (TRU). The objectives of the study are: to determine if removal of radionuclides from HLW and TRU significantly reduces the long-term radiological risks of geologic disposal; to determine if chemical partitioning of the waste for space disposal is technically feasible; to identify acceptable waste forms for space disposal; and to compare improvements in geologic disposal system performance to impacts of additional treatment, storage, and transportation necessary for space disposal. To compare radiological effects, five system alternatives are defined: Reference case - All HLW and TRU to a repository. Alternative A - Iodine to space, the balance to a repository. Alternative B - Technetium to space, the balance to a repository. Alternative C - 95% of cesium and strontium to a repository; the balance of HLW aged first, then to space; plutonium separated from TRU for recycle; the balance of the TRU to a repository. Alternative D - HLW aged first, then to space, plutonium separated from TRU for recycle; the balance of the TRU to a repository. The conclusions of this study are: the incentive for space disposal is that it offers a perception of reduced risks rather than significant reduction. Suitable waste forms for space disposal are cermet for HLW, metallic technetium, and lead iodide. Space disposal of HLW appears to offer insignificant safety enhancements when compared to geologic disposal; the disposal of iodine and technetium wastes in space does not offer risk advantages. Increases in short-term doses for the alternatives are minimal; however, incremental costs of treating, storing and transporting wastes for space disposal are substantial

  12. Effects of feed process variables on Hanford Vitrification Plant performance

    International Nuclear Information System (INIS)

    Farnsworth, R.K.; Peterson, M.E.; Wagner, R.N.

    1987-01-01

    As a result of nuclear defense activities, high-level liquid radioactive wastes have been generated at the Hanford Site for over 40 yr. The Hanford Waste Vitrification Plant (HWVP) is being proposed to immobilize these wastes in a waste form suitable for disposal in a geologic repository. Prior to vitrification, the waste will undergo several conditioning steps before being fed to the melter. The effect of certain process variables on the resultant waste slurry properties must be known to assure processability of the waste slurry during feed preparation. Of particular interest are the rheological properties, which include the yield stress and apparent viscosity. Identification of the rheological properties of the slurry is required to adequately design the process equipment used for feed preparation (agitators, mixing tanks, concentrators, etc.). Knowledge of the slurry rheological properties is also necessary to establish processing conditions and operational limits for maximum plant efficiency and reliability. A multivariable study was performed on simulated HWVP feed to identify the feed process variables that have a significant impact on rheology during processing. Two process variables were evaluated in this study: (a) the amount of formic acid added to the feed and (b) the degree of shear encountered by the feed during processing. The feed was physically and rheologically characterized at various stages during feed processing

  13. National Institutes of Health: Mixed waste stream analysis

    International Nuclear Information System (INIS)

    Kirner, N.P.; Faison, G.P.; Johnson, D.R.

    1994-08-01

    The Low-Level Radioactive Waste Policy Amendments Act of 1985 requires that the US Department of Energy (DOE) provide technical assistance to host States, compact regions, and unaffiliated States to fulfill their responsibilities under the Act. The National Low-Level Waste Management Program (NLLWMP) operated for DOE by EG ampersand G Idaho, Inc. provides technical assistance in the development of new commercial low-level radioactive waste disposal capacity. The NLLWMP has been requested by the Appalachian Compact to help the biomedical community become better acquainted with its mixed waste streams, to help minimize the mixed waste streams generated by the biomedical community, and to provide applicable treatment technologies to those particular mixed waste streams. Mixed waste is waste that satisfies the definition of low-level radioactive waste (LLW) in the Low-Level Radioactive Waste Policy Act of 1980 (LLRWPA) and contains hazardous waste that either (a) is listed as a hazardous waste in Subpart D of 40 CFR 261, or (b) causes the LLW to exhibit any of the hazardous waste characteristics identified in 40 CFR 261. The purpose of this report is to clearly define and characterize the mixed waste streams generated by the biomedical community so that an identification can be made of the waste streams that can and cannot be minimized and treated by current options. An understanding of the processes and complexities of generation of mixed waste in the biomedical community may encourage more treatment and storage options to become available

  14. Mixed waste and waste minimization: The effect of regulations and waste minimization on the laboratory

    International Nuclear Information System (INIS)

    Dagan, E.B.; Selby, K.B.

    1993-08-01

    The Hanford Site is located in the State of Washington and is subject to state and federal environmental regulations that hamper waste minimization efforts. This paper addresses the negative effect of these regulations on waste minimization and mixed waste issues related to the Hanford Site. Also, issues are addressed concerning the regulations becoming more lenient. In addition to field operations, the Hanford Site is home to the Pacific Northwest Laboratory which has many ongoing waste minimization activities of particular interest to laboratories

  15. A study on safety assessment methodology for a vitrification plant

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Y. C.; Lee, G. S.; Choi, Y. C.; Kim, G. H. [Yonsei Univ., Seoul (Korea, Republic of)

    2002-03-15

    In this study, the technical and regulatory status of radioactive waste vitrification technologies in foreign and domestic plants is investigated and analyzed, and then significant factors are suggested which must be contained in the final technical guideline or standard for the safety assessment of vitrification plants. Also, the methods to estimate the stability of vitrified waste forms are suggested with property analysis of them. The contents and scope of the study are summarized as follows : survey of the status on radioactive waste vitrification technologies in foreign and domestic plants, survey of the characterization methodology for radioactive waste form, analysis of stability for vitrified waste forms, survey and analysis of technical standards and regulations concerned with them in foreign and domestic plants, suggestion of significant factors for the safety assessment of vitrification plants, submission of regulated technical standard on radioactive waste vitrification plats.

  16. Alternative oxidation technologies for organic mixed waste

    International Nuclear Information System (INIS)

    Borduin, L.C.; Fewell, T.

    1998-01-01

    The Mixed Waste Focus Area (MWFA) is currently supporting the development and demonstration of several alternative oxidation technology (AOT) processes for treatment of combustible mixed low-level wastes. AOTs have been defined as technologies that destroy organic material without using open-flame reactions. AOTs include both thermal and nonthermal processes that oxidize organic wastes but operate under significantly different physical and chemical conditions than incinerators. Nonthermal processes currently being studied include Delphi DETOX and acid digestion at the Savannah River Site (SRS), and direct chemical oxidation at Lawrence Livermore National Laboratory (LLNL). All three technologies are at advanced stages of development or are entering the demonstration phase. Nonflame thermal processes include catalytic chemical oxidation, which is being developed and deployed at Lawrence Berkeley National Laboratory (LBNL), and steam reforming, a commercial process being supported by the Department of Energy (DOE). Although testing is complete on some AOT technologies, most require additional support to complete some or all of the identified development objectives. Brief descriptions, status, and planned paths forward for each of the technologies are presented

  17. Remediating the INEL's buried mixed waste tanks

    International Nuclear Information System (INIS)

    Kuhns, D.J.; Matthern, G.E.; Reese, C.L.

    1996-01-01

    The Idaho National Engineering Laboratory (INEL), formerly the National Reactor Testing Station (NRTS), encompasses 890 square miles and is located in southeast Idaho. In 1949, the United States Atomic Energy Commission, now the Department of Energy (DOE), established the NRTS as a site for the building and testing of nuclear facilities. Wastes generated during the building and testing of these nuclear facilities were disposed within the boundaries of the site. These mixed wastes, containing radionuclides and hazardous materials, were often stored in underground tanks for future disposal. The INEL has 11 buried mixed waste storage tanks regulated under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) ranging in size from 400 to 50,000 gallons. These tanks are constructed of either stainless or carbon steel and are located at 3 distinct geographic locations across the INEL. These tanks have been grouped based on their similarities in an effort to save money and decrease the time required to complete the necessary remediation. Environmental Restoration and Technology Development personnel are teaming in an effort to address the remediation problem systematically

  18. Evaluation of vitrification factors from DWPF's macro-batch 1

    International Nuclear Information System (INIS)

    Edwards, T.B.

    2000-01-01

    The Defense Waste Processing Facility (DWPF) is evaluating new sampling and analytical methods that may be used to support future Slurry Mix Evaporator (SME) batch acceptability decisions. This report uses data acquired during DWPF's processing of macro-batch 1 to determine a set of vitrification factors covering several SME and Melter Feed Tank (MFT) batches. Such values are needed for converting the cation measurements derived from the new methods to a ''glass'' basis. The available data from macro-batch 1 were used to examine the stability of these vitrification factors, to estimate their uncertainty over the course of a macro-batch, and to provide a recommendation on the use of a single factor for an entire macro-batch. The report is in response to Technical Task Request HLW/DWPF/TTR-980015

  19. Mixed Waste Management Options: 1995 Update. National Low-Level Waste Management Program

    Energy Technology Data Exchange (ETDEWEB)

    Kirner, N.; Kelly, J.; Faison, G.; Johnson, D. [Foster Wheeler Environmental Corp. (United States)

    1995-05-01

    In the original mixed Waste Management Options (DOE/LLW-134) issued in December 1991, the question was posed, ``Can mixed waste be managed out of existence?`` That study found that most, but not all, of the Nation`s mixed waste can theoretically be managed out of existence. Four years later, the Nation is still faced with a lack of disposal options for commercially generated mixed waste. However, since publication of the original Mixed Waste Management Options report in 1991, limited disposal capacity and new technologies to treat mixed waste have become available. A more detailed estimate of the Nation`s mixed waste also became available when the US Environmental Protection Agency (EPA) and the US Nuclear Regulatory Commission (NRC) published their comprehensive assessment, titled National Profile on Commercially Generated Low-Level Radioactive Mixed Waste (National Profile). These advancements in our knowledge about mixed waste inventories and generation, coupled with greater treatment and disposal options, lead to a more applied question posed for this updated report: ``Which mixed waste has no treatment option?`` Beyond estimating the volume of mixed waste requiring jointly regulated disposal, this report also provides a general background on the Atomic Energy Act (AEA) and the Resource Conservation and Recovery Act (RCRA). It also presents a methodical approach for generators to use when deciding how to manage their mixed waste. The volume of mixed waste that may require land disposal in a jointly regulated facility each year was estimated through the application of this methodology.

  20. Mixed Waste Management Options: 1995 Update. National Low-Level Waste Management Program

    International Nuclear Information System (INIS)

    Kirner, N.; Kelly, J.; Faison, G.; Johnson, D.

    1995-05-01

    In the original mixed Waste Management Options (DOE/LLW-134) issued in December 1991, the question was posed, ''Can mixed waste be managed out of existence?'' That study found that most, but not all, of the Nation's mixed waste can theoretically be managed out of existence. Four years later, the Nation is still faced with a lack of disposal options for commercially generated mixed waste. However, since publication of the original Mixed Waste Management Options report in 1991, limited disposal capacity and new technologies to treat mixed waste have become available. A more detailed estimate of the Nation's mixed waste also became available when the US Environmental Protection Agency (EPA) and the US Nuclear Regulatory Commission (NRC) published their comprehensive assessment, titled National Profile on Commercially Generated Low-Level Radioactive Mixed Waste (National Profile). These advancements in our knowledge about mixed waste inventories and generation, coupled with greater treatment and disposal options, lead to a more applied question posed for this updated report: ''Which mixed waste has no treatment option?'' Beyond estimating the volume of mixed waste requiring jointly regulated disposal, this report also provides a general background on the Atomic Energy Act (AEA) and the Resource Conservation and Recovery Act (RCRA). It also presents a methodical approach for generators to use when deciding how to manage their mixed waste. The volume of mixed waste that may require land disposal in a jointly regulated facility each year was estimated through the application of this methodology

  1. In Situ Vitrification Engineering-Scale Test ES-INEL-4 Product Characterization Test Plan

    International Nuclear Information System (INIS)

    Weidner, J.R.; Stoots, P.R.

    1990-06-01

    In 1987, the Buried Waste Program (BWP) was established within EG ampersand G Idaho, Inc., the prime contractor at INEL. Following the Environmental Restoration guidelines of the Buried Waste Program, the In Situ Vitrification Program is participating in a Remedial Investigation/Feasibility Study (RI/FS) for permanent disposal of INEL waste, in compliance with the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). This study was requested and is being funded by the Office of Technology Development of the Idaho Operations Office of DOE (DOE-ID). As part of the RI/FS, an in situ vitrification (ISV) scoping study on the treatability of mixed low-level and mixed transuranic-contaminated waste is being performed to determine the applicability of ISV to remediation of waste at SDA. In examination of the ISV process for applicability to SDA waste, this In Situ Vitrification Engineering-Scale Test ES-INEL-4 Product Characterization Test Plan identifies the following: sampling and analysis strategy; sampling procedures; methods to conduct analyses; equipment; and procedures to ensure data quality. 8 refs., 2 tabs

  2. Bioprocessing of concentrated mixed hazardous industrial waste

    International Nuclear Information System (INIS)

    Wolfram, J.H.; Rogers, R.D.; Silver, G.; Attalla, A.; Prisc, M.

    1994-01-01

    The use of selected microorganisms for the degradation and/or the detoxification of hazardous organic compounds is gaining wide acceptance as an alternative waste treatment technology. This work describes the unique capabilities of an isolated strain of Pseudomonas for metabolizing methylated aromatic compounds. This strain of Pseudomonas putida Idaho is unique in that it can tolerate and grow under a layer of neat p-xylene. A bioprocess has been developed to degrade LLW and mixed wastes containing methylated aromatic compounds, i.e., pseudocumene, toluene and p-xylene. The process is now in the demonstration phase at a DOE facility and has been running for one year. Feed concentrations of 21200 ppm of the toxic organic substrate have been fed to the bioreactor. This report describes the results obtained thus far

  3. DOE acceptance of commercial mixed waste -- Studies are under way

    Energy Technology Data Exchange (ETDEWEB)

    Plummer, T.L. [Dept. of Energy, Washington, DC (United States). Technical Support Program; Owens, C.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States). National Low-Level Waste Management Program

    1993-03-01

    The topic of the Department of Energy acceptance of commercial mixed waste at DOE facilities has been proposed by host States and compact regions that are developing low-level radioactive waste disposal facilities. States support the idea of DOE accepting commercial mixed waste because (a) very little commercial mixed waste is generated compared to generation by DOE facilities (Department of Energy--26,300 cubic meters annually vs. commercial--3400 cubic meters annually); (b) estimated costs for commercial disposal are estimated to be $15,000 to $40,000 per cubic foot; (c) once treatment capability becomes available, 70% of the current levels of commercial mixed waste will be eliminated, (d) some State laws prohibit the development of mixed waste disposal facilities in their States; (e) DOE is developing a nationwide strategy that will include treatment and disposal capacity for its own mixed waste and the incremental burden on the DOE facilities would be minuscule, and (6) no States are developing mixed waste disposal facilities. DOE senior management has repeatedly expressed willingness to consider investigating the feasibility of DOE accepting commercial mixed waste. In January 1991, Leo Duffy of the Department of energy met with members of the Low-Level Radioactive Waste Forum, which led to an agreement to explore such an arrangement. He stated that this seems like a cost-effective way to solve commercial mixed waste management problems.

  4. Mixed and Low-Level Treatment Facility Project. Appendix B, Waste stream engineering files, Part 1, Mixed waste streams

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This appendix contains the mixed and low-level waste engineering design files (EDFS) documenting each low-level and mixed waste stream investigated during preengineering studies for Mixed and Low-Level Waste Treatment Facility Project. The EDFs provide background information on mixed and low-level waste generated at the Idaho National Engineering Laboratory. They identify, characterize, and provide treatment strategies for the waste streams. Mixed waste is waste containing both radioactive and hazardous components as defined by the Atomic Energy Act and the Resource Conservation and Recovery Act, respectively. Low-level waste is waste that contains radioactivity and is not classified as high-level waste, transuranic waste, spent nuclear fuel, or 11e(2) byproduct material as defined by DOE 5820.2A. Test specimens of fissionable material irradiated for research and development only, and not for the production of power or plutonium, may be classified as low-level waste, provided the concentration of transuranic is less than 100 nCi/g. This appendix is a tool that clarifies presentation format for the EDFS. The EDFs contain waste stream characterization data and potential treatment strategies that will facilitate system tradeoff studies and conceptual design development. A total of 43 mixed waste and 55 low-level waste EDFs are provided.

  5. Mixed waste removal from a hazardous waste storage tank

    International Nuclear Information System (INIS)

    Geber, K.R.

    1993-01-01

    The spent fuel transfer canal at the Oak Ridge Graphite Reactor was found to be leaking 400 gallons of water per day into the surrounding soil. Sampling of the sediment layer on the floor of the canal to determine the environmental impact of the leak identified significant radiological contamination and elevated levels of cadmium and lead which are hazardous under the Resource Conservation and Recovery Act (RCRA). Under RCRA regulations and Rules of Tennessee Department of Environment and Conservation, the canal was considered a hazardous waste storage tank. This paper describes elements of the radiological control program established in support of a fast-track RCRA closure plan that involved underwater mapping of the radiation fields, vacuuming, and ultra-filtration techniques that were successfully used to remove the mixed waste sediments and close the canal in a method compliant with state and federal regulations

  6. Is radioactive mixed waste packaging and transportation really a problem

    International Nuclear Information System (INIS)

    McCall, D.L.; Calihan, T.W. III.

    1992-01-01

    Recently, there has been significant concern expressed in the nuclear community over the packaging and transportation of radioactive mixed waste under US Department of Transportation regulation. This concern has grown more intense over the last 5 to 10 years. Generators and regulators have realized that much of the waste shipped as ''low-level radioactive waste'' was in fact ''radioactive mixed waste'' and that these wastes pose unique transportation and disposal problems. Radioactive mixed wastes must, therefore, be correctly identified and classed for shipment. If must also be packaged, marked, labeled, and otherwise prepared to ensure safe transportation and meet applicable storage and disposal requirements, when established. This paper discusses regulations applicable to the packaging and transportation of radioactive mixed waste and identifies effective methods that waste shippers can adopt to meet the current transportation requirements. This paper will include a characterization and description of the waste, authorized packaging, and hazard communication requirements during transportation. Case studies will be sued to assist generators in understanding mixed waste shipment requirements and clarify the requirements necessary to establish a waste shipment program. Although management and disposal of radioactive mixed waste is clearly a critical issue, packaging and transportation of these waste materials is well defined in existing US Department of Transportation hazardous material regulations

  7. Development of Characterization Protocol for Mixed Liquid Radioactive Waste Classification

    International Nuclear Information System (INIS)

    Norasalwa Zakaria; Syed Asraf Wafa; Wo, Y.M.; Sarimah Mahat; Mohamad Annuar Assadat Husain

    2017-01-01

    Mixed organic liquid waste generated from health-care and research activities containing tritium, carbon-14, and other radionuclide posed specific challenges in its management. Often, this waste becomes legacy waste in many nuclear facilities and being considered as 'problematic' waste. One of the most important recommendations made by IAEA is to perform multistage processes aiming at declassification of the waste. At this moment, approximately 3000 bottles of mixed liquid waste, with estimated volume of 6000 litres are currently stored at the National Radioactive Waste Management Centre, Malaysia and some have been stored for more than 25 years. The aim of this study is to develop a characterization protocol towards reclassification of these wastes. The characterization protocol entails waste identification, waste screening and segregation, and analytical radionuclides profiling using analytical procedures involving gross alpha beta, and gamma spectrometry. The results obtained from the characterization protocol are used to establish criteria for speedy classification of the waste. (author)

  8. Development of characterization protocol for mixed liquid radioactive waste classification

    Energy Technology Data Exchange (ETDEWEB)

    Zakaria, Norasalwa, E-mail: norasalwa@nuclearmalaysia.gov.my [Waste Technology Development Centre, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia); Wafa, Syed Asraf [Radioisotop Technology and Innovation, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia); Wo, Yii Mei [Radiochemistry and Environment, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia); Mahat, Sarimah [Material Technology Group, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Mixed liquid organic waste generated from health-care and research activities containing tritium, carbon-14, and other radionuclides posed specific challenges in its management. Often, these wastes become legacy waste in many nuclear facilities and being considered as ‘problematic’ waste. One of the most important recommendations made by IAEA is to perform multistage processes aiming at declassification of the waste. At this moment, approximately 3000 bottles of mixed liquid waste, with estimated volume of 6000 litres are currently stored at the National Radioactive Waste Management Centre, Malaysia and some have been stored for more than 25 years. The aim of this study is to develop a characterization protocol towards reclassification of these wastes. The characterization protocol entails waste identification, waste screening and segregation, and analytical radionuclides profiling using various analytical procedures including gross alpha/ gross beta, gamma spectrometry, and LSC method. The results obtained from the characterization protocol are used to establish criteria for speedy classification of the waste.

  9. Summary of BNL studies regarding commercial mixed waste

    International Nuclear Information System (INIS)

    Bowerman, B.S.; Kempf, C.R.; MacKenzie, D.R.; Siskind, B.; Piciulo, P.L.

    1986-09-01

    Based on BNL's study it was concluded that there are low-level radioactive wastes (LLWs) which contain chemically hazardous components. Scintillation liquids may be considered an EPA listed hazardous waste and are, therefore, potential mixed wastes. Since November 1985, no operating LLW disposal site will accept these wastes for disposal. Unless such wastes contain de minimis quantities of radionuclides, they cannot be disposed of at an EPA permitted site. Currently generators of liquid scintillation wastes can ship de minimis wastes to be burned at commercial facilities. Oil wastes may also eventually be an EPA listed waste and thus will have to be considered a potential radioactive mixed waste unless NRC establishes de minimis levels of radionuclides below which oils can be managed as hazardous wastes. Regarding wastes containing lead metal there is some question as to the extent of the hazard posed by lead disposed in a LLW burial trench. Chromium-containing wastes would have to be tested to determine whether they are potential mixed wastes. There may be other wastes that are mixed wastes; the responsibility for determining this rests with the waste generator. While management options for handling potential mixed wastes are available, there is limited regulatory guidance for generators. BNL has identified and evaluated a variety of treatment options for the management of potential radioactive mixed wastes. The findings of that study showed that application of a management option with the purpose of addressing EPA concerns can, at the same time, address stabilization and volume reduction concerns of NRC. 6 refs., 1 tab

  10. Design and performance of a 100-kg/h, direct calcine-fed electric-melter system for nuclear-waste vitrification

    International Nuclear Information System (INIS)

    Dierks, R.D.

    1980-11-01

    This report describes the physical characteristics of a ceramic-lined, joule-heated glass melter that is directly connected to the discharge of a spray calciner and is currently being used to study the vitrification of simulated nuclear-waste slurries. Melter performance characteristics and subsequent design improvements are described. The melter contains 0.24 m 3 of glass with a glass surface area of 0.76 m 2 , and is heated by the flow of an alternating current (ranging from 600 to 1200 amps) between two Inconel-690 slab-type electrodes immersed in the glass at either end of the melter tank. The melter was maintained at operating temperature (900 to 1260 0 C) for 15 months, and produced 62,000 kg of glass. The maximum sustained operating period was 122 h, during which glass was produced at the rate of 70 kg/h

  11. DOE looks to clean up with vitrification technology

    International Nuclear Information System (INIS)

    Lobsenz, G.

    1994-01-01

    This article describes the vitrification and waste retrieval facility being built by US DOE, designed to handle a mixture of low-level radioactive wastes stored in structurally shaky silos at the Fernald weapons plant

  12. Surrogate formulations for thermal treatment of low-level mixed waste, Part II: Selected mixed waste treatment project waste streams

    Energy Technology Data Exchange (ETDEWEB)

    Bostick, W.D.; Hoffmann, D.P.; Chiang, J.M.; Hermes, W.H.; Gibson, L.V. Jr.; Richmond, A.A. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States); Mayberry, J. [Science Applications International Corp., Idaho Falls, ID (United States); Frazier, G. [Univ. of Tennessee, Knoxville, TN (United States)

    1994-01-01

    This report summarizes the formulation of surrogate waste packages, representing the major bulk constituent compositions for 12 waste stream classifications selected by the US DOE Mixed Waste Treatment Program. These waste groupings include: neutral aqueous wastes; aqueous halogenated organic liquids; ash; high organic content sludges; adsorbed aqueous and organic liquids; cement sludges, ashes, and solids; chloride; sulfate, and nitrate salts; organic matrix solids; heterogeneous debris; bulk combustibles; lab packs; and lead shapes. Insofar as possible, formulation of surrogate waste packages are referenced to authentic wastes in inventory within the DOE; however, the surrogate waste packages are intended to represent generic treatability group compositions. The intent is to specify a nonradiological synthetic mixture, with a minimal number of readily available components, that can be used to represent the significant challenges anticipated for treatment of the specified waste class. Performance testing and evaluation with use of a consistent series of surrogate wastes will provide a means for the initial assessment (and intercomparability) of candidate treatment technology applicability and performance. Originally the surrogate wastes were intended for use with emerging thermal treatment systems, but use may be extended to select nonthermal systems as well.

  13. Surrogate formulations for thermal treatment of low-level mixed waste, Part II: Selected mixed waste treatment project waste streams

    International Nuclear Information System (INIS)

    Bostick, W.D.; Hoffmann, D.P.; Chiang, J.M.; Hermes, W.H.; Gibson, L.V. Jr.; Richmond, A.A.; Mayberry, J.; Frazier, G.

    1994-01-01

    This report summarizes the formulation of surrogate waste packages, representing the major bulk constituent compositions for 12 waste stream classifications selected by the US DOE Mixed Waste Treatment Program. These waste groupings include: neutral aqueous wastes; aqueous halogenated organic liquids; ash; high organic content sludges; adsorbed aqueous and organic liquids; cement sludges, ashes, and solids; chloride; sulfate, and nitrate salts; organic matrix solids; heterogeneous debris; bulk combustibles; lab packs; and lead shapes. Insofar as possible, formulation of surrogate waste packages are referenced to authentic wastes in inventory within the DOE; however, the surrogate waste packages are intended to represent generic treatability group compositions. The intent is to specify a nonradiological synthetic mixture, with a minimal number of readily available components, that can be used to represent the significant challenges anticipated for treatment of the specified waste class. Performance testing and evaluation with use of a consistent series of surrogate wastes will provide a means for the initial assessment (and intercomparability) of candidate treatment technology applicability and performance. Originally the surrogate wastes were intended for use with emerging thermal treatment systems, but use may be extended to select nonthermal systems as well

  14. Kinetics of Cold-Cap Reactions for Vitrification of Nuclear Waste Glass Based on Simultaneous Differential Scanning Calorimetry - Thermogravimetry (DSC-TGA) and Evolved Gas Analysis (EGA)

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, Carmen P.; Pierce, David A.; Schweiger, Michael J.; Kruger, Albert A.; Chun, Jaehun; Hrma, Pavel R.

    2013-12-03

    For vitrifying nuclear waste glass, the feed, a mixture of waste with glass-forming and modifying additives, is charged onto the cold cap that covers 90-100% of the melt surface. The cold cap consists of a layer of reacting molten glass floating on the surface of the melt in an all-electric, continuous glass melter. As the feed moves through the cold cap, it undergoes chemical reactions and phase transitions through which it is converted to molten glass that moves from the cold cap into the melt pool. The process involves a series of reactions that generate multiple gases and subsequent mass loss and foaming significantly influence the mass and heat transfers. The rate of glass melting, which is greatly influenced by mass and heat transfers, affects the vitrification process and the efficiency of the immobilization of nuclear waste. We studied the cold-cap reactions of a representative waste glass feed using both the simultaneous differential scanning calorimetry thermogravimetry (DSC-TGA) and the thermogravimetry coupled with gas chromatography-mass spectrometer (TGA-GC-MS) as complementary tools to perform evolved gas analysis (EGA). Analyses from DSC-TGA and EGA on the cold-cap reactions provide a key element for the development of an advanced cold-cap model. It also helps to formulate melter feeds for higher production rate.