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Sample records for waste vitrification plan

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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.

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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.)

  6. 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

  7. 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)

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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)

  2. 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

  3. 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.

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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.)

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. In situ vitrification engineering-scale test ES-INEL-5 test plan

    International Nuclear Information System (INIS)

    Stoots, P.R.

    1990-06-01

    In 1952, the Radioactive Waste Management Complex (RWMC) was established at the Idaho National Engineering Laboratory (INEL). RWMC is located on approximately 144 acres in the southwestern corner of the INEL site and was established as a controlled area for the burial of solid low-level wastes generated by INEL operations. In 1954, the 88-acre Subsurface Disposal Area (SDA) of RWMC began accepting solid transuranic-contaminated waste. From 1954 to 1970, transuranic-contaminated waste was accepted from the Rocky Flats Plant (RFP) near Golden, CO, as well as from other US Department of Energy (DOE) locations. In 1987, the Buried Waste Program (BWP) was established by 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 applicability of ISV to remediation of waste at SDA. This In Situ Vitrification Engineering-Scale Test ES-INEL-5 Test Plan considers the data needs of engineering, regulatory, health, and safety activities for all sampling and analysis activities in support of engineering scale test ES-INEL-5. 5 refs., 3 figs., 4 tabs

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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.

  7. 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

  8. Vitrification development and experiences at Fernald, Ohio

    International Nuclear Information System (INIS)

    Gimpel, R.F.; Paine, D.; Roberts, J.L.; Akgunduz, N.

    1998-01-01

    Vitrification of radioactive wastes products have proven to produce an extremely stable waste form. Vitrification involves the melting of wastes with a mixture of glass-forming additives at high temperatures; when cooled, the wastes are incorporated into a glass that is analogous to obsidian. Obsidian is a volcanic glass-like rock, commonly found in nature. A one-metric ton/day Vitrification Pilot Plant (VITPP) at Fernald, Ohio, simulated the vitrification of radium and radon bearing silo residues using representative non-radioactive surrogates. These non-radioactive surrogates contained high concentrations of lead, sulfates, and phosphates. The vitrification process was carried out at temperatures of 1150 to 1350 C. Laboratory and bench-scale treatability studies were conducted before initiation of the VITPP. Development of the glass formulas, containing up to 90% waste, will be discussed in the paper. The VITPP processed glass for seven months, until a breach of the melter containment vessel suspended operations. More than 70,000 pounds of good surrogate glass were produced by the VITPP. Experiences, lessons learned, and the planned path forward will be presented

  9. 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.

  10. 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

  11. 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.

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. Interim Hanford Waste Management Technology Plan

    International Nuclear Information System (INIS)

    1985-09-01

    The Interim Hanford Waste Management Technology Plan (HWMTP) is a companion document to the Interim Hanford Waste Management Plan (HWMP). A reference plan for management and disposal of all existing and certain projected future radioactive Hanford Site Defense Wastes (HSDW) is described and discussed in the HWMP. Implementation of the reference plan requires that various open technical issues be satisfactorily resolved. The principal purpose of the HWMTP is to present detailed descriptions of the technology which must be developed to close each of the technical issues associated with the reference plan identified in the HWMP. If alternative plans are followed, however, technology development efforts including costs and schedules must be changed accordingly. Technical issues addressed in the HWMTP and HWMP are those which relate to disposal of single-shell tank wastes, contaminated soil sites, solid waste burial sites, double-shell tank wastes, encapsulated 137 CsCl and 90 SrF 2 , stored and new solid transuranic (TRU) wastes, and miscellaneous wastes such as contaminated sodium metal. Among the high priority issues to be resolved are characterization of various wastes including early determination of the TRU content of future cladding removal wastes; completion of development of vitrification (Hanford Waste Vitrification Plant) and grout technology; control of subsidence in buried waste sites; and development of criteria and standards including performance assessments of systems proposed for disposal of HSDW. Estimates of the technology costs shown in this report are made on the basis that all identified tasks for all issues associated with the reference disposal plan must be performed. Elimination of, consolidation of, or reduction in the scope of individual tasks will, of course, be reflected in corresponding reduction of overall technology costs

  19. Tank Waste Remediation System Projects Document Control Plan

    International Nuclear Information System (INIS)

    Slater, G.D.; Halverson, T.G.

    1994-01-01

    The purpose of this Tank Waste Remediation System Projects Document Control Plan is to provide requirements and responsibilities for document control for the Hanford Waste Vitrification Plant (HWVP) Project and the Initial Pretreatment Module (IPM) Project

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. 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.

  17. 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

    Vitrification is the technology that has been chosen to solidify approximately 15,500 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. 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. The K-65 residues contain radium, uranium, uranium daughter products, and heavy metals such as lead and barium.The K-65 waste leaches lead at greater than 100 times the allowable Environmental Protection Agency (EPA) Resource, Conservation, and Recovery Act (RCRA) concentration limits when tested by the Toxic Characteristic Leaching Procedure (TCLP). Vitrification was chosen by FEMP as the preferred technology for the Silos 1, 2, 3 wastes because the final waste form met the following criteria: controls radon emanation, eliminates the potential for hazardous or radioactive constituents to migrate to the aquifer below FEMP, controls the spread of radioactive particulates, reduces leachability of metals and radiological constituents, reduces volume of final wasteform for disposal, silo waste composition is favorable to vitrification, will meet current and proposed RCRA TCLP leaching criteria Glasses that melt at 1350 degrees C were developed by Pacific Northwest National Laboratory (PNNL) and glasses that melt between 1150-1350 degrees C were developed by the Vitreous State Laboratory (VSL) for the K-65 silo wastes. Both crucible studies and pilot scale vitrification studies were conducted by PNNL and VSL. Subsequently, a Vitrification Pilot Plant (VPP) was constructed at FEMP capable of operating at temperatures up to 1450

  18. 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)

  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. 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

  1. 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

  2. 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

  3. 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.

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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

  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 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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.

  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. 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

  4. 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)

  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. 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.

  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. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. Vitrification of HLW in cold crucible melter

    International Nuclear Information System (INIS)

    Bordier, G.

    2005-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel the CEA (French Atomic Energy Commission), COGEMA (Industrial Operator), and SGN (COGEMA's Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities: the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. Secondary Waste Cast Stone Waste Form Qualification Testing Plan

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H.; Serne, R. Jeffrey

    2012-09-26

    The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). Cast Stone – a cementitious waste form, has been selected for solidification of this secondary waste stream after treatment in the ETF. The secondary-waste Cast Stone waste form must be acceptable for disposal in the IDF. This secondary waste Cast Stone waste form qualification testing plan outlines the testing of the waste form and immobilization process to demonstrate that the Cast Stone waste form can comply with the disposal requirements. Specifications for the secondary-waste Cast Stone waste form have not been established. For this testing plan, Cast Stone specifications are derived from specifications for the immobilized LAW glass in the WTP contract, the waste acceptance criteria for the IDF, and the waste acceptance criteria in the IDF Permit issued by the State of Washington. This testing plan outlines the testing needed to demonstrate that the waste form can comply with these waste form specifications and acceptance criteria. The testing program must also demonstrate that the immobilization process can be controlled to consistently provide an acceptable waste form product. This testing plan also outlines the testing needed to provide the technical basis for understanding the long-term performance of the waste form in the disposal environment. These waste form performance data are needed to support performance assessment analyses of the long-term environmental impact of the secondary-waste Cast Stone waste form in the IDF

  8. 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.

  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. 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

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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.

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. Vitrification facility at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    DesCamp, V.A.; McMahon, C.L.

    1996-07-01

    This report is a description of the West Valley Demonstration Project's vitrification facilities from the establishment of the West Valley, NY site as a federal and state cooperative project to the completion of all activities necessary to begin solidification of radioactive waste into glass by vitrification. Topics discussed in this report include the Project's background, high-level radioactive waste consolidation, vitrification process and component testing, facilities design and construction, waste/glass recipe development, integrated facility testing, and readiness activities for radioactive waste processing

  11. 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)

  12. 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

  13. 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

  14. 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

  15. Test Plan: Phase 1 demonstration of 3-phase electric arc melting furnace technology for vitrifying high-sodium content low-level radioactive liquid wastes

    Energy Technology Data Exchange (ETDEWEB)

    Eaton, W.C. [ed.

    1995-05-31

    This document provides a test plan for the conduct of electric 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-384216] is the US Bureau of Mines, Department of the Interior, Albany Research Center, Albany, Oregon. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes feed preparation activities and melting of glass with Hanford LLW Double-Shell Slurry Feed waste simulant in a 3-phase electric arc (carbon electrode) furnace.

  16. Test Plan: Phase 1 demonstration of 3-phase electric arc melting furnace technology for vitrifying high-sodium content low-level radioactive liquid wastes

    International Nuclear Information System (INIS)

    Eaton, W.C.

    1995-01-01

    This document provides a test plan for the conduct of electric 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-384216] is the US Bureau of Mines, Department of the Interior, Albany Research Center, Albany, Oregon. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes feed preparation activities and melting of glass with Hanford LLW Double-Shell Slurry Feed waste simulant in a 3-phase electric arc (carbon electrode) furnace

  17. 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

  18. 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.

  19. 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

  20. 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.%.

  1. 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

  2. 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)

  3. 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)

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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...

  14. 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

  15. 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

  16. 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.'

  17. 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

  18. 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

  19. 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.

  20. 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.

  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. 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

  3. 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)

  4. 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.

  5. 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.

  6. 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

  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. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  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. 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)

  16. Rocky Flats Plant precipitate sludge surrogate vitrification demonstration. Technical Task Plan

    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. The preferred disposal method would be one that is capable of consistently producing a durable leach resistant wasteform, while simultaneously minimizing disposal volumes. 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

  17. Test Plan: Phase 1, Hanford LLW melter tests, GTS Duratek, Inc

    International Nuclear Information System (INIS)

    Eaton, W.C.

    1995-01-01

    This document provides a test plan for the conduct of 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-384215] is GTS Duratek, Inc., Columbia, Maryland. The GTS Duratek project manager for this work is J. Ruller. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes melting of glass with Hanford LLW Double-Shell Slurry Feed waste simulant in a DuraMelter trademark vitrification system

  18. PNL vitrification technology development project glass formulation strategy for LLW vitrification

    International Nuclear Information System (INIS)

    Kim, D.; Hrma, P.R.; Westsik, J.H. Jr.

    1996-03-01

    This Glass Formulation Strategy describes development approaches to optimize glass compositions for Hanford's low-level waste vitrification between now and the projected low-level waste facility start-up in 2005. The objectives of the glass formulation task are to develop optimized glass compositions with satisfactory long-term durability, acceptable processing characteristics, adequate flexibility to handle waste variations, maximize waste loading to practical limits, and to develop methodology to respond to further waste variations

  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. 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

  1. Hanford low-level vitrification melter testing -- Master list of data submittals

    International Nuclear Information System (INIS)

    Hendrickson, D.W.

    1995-01-01

    The Westinghouse Hanford Company (WHC) is conducting a two-phased effort to evaluate melter system technologies for vitrification of liquid low-level radioactive waste (LLW) streams. The evaluation effort includes demonstration testing of selected glass melter technologies and technical reports regarding the applicability of the glass melter technologies to the vitrification of Hanford LLW tank waste. The scope of this document is to identify and list vendor document submittals in technology demonstration support of the Hanford Low-Level Waste Vitrification melter testing program. The scope of this document is limited to those documents responsive to the Statement of Work, accepted and issued by the LLW Vitrification Program. The purpose of such a list is to maintain configuration control of vendor supplied data and to enable ready access to, and application of, vendor supplied data in the evaluation of melter technologies for the vitrification of Hanford low-level tank wastes

  2. 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

  3. 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.

  4. 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

  5. Vitrification of HLW produced by uranium/molybdenum fuel reprocessing in cogema's cold crucible melter

    International Nuclear Information System (INIS)

    Quang, R. Do; Petitjean, V.; Hollebeque, F.; Pinet, O.; Flament, T.; Prodhomme, A.; Dalcorso, J. P.

    2003-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R and D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed

  6. Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter

    International Nuclear Information System (INIS)

    Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod'homme, A.

    2003-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R and D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed

  7. 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.

  8. Vitrification of organic products in a cold crucible

    International Nuclear Information System (INIS)

    Song, Myung Jae; Park, Jong Kil; Jouan, A.; Ladirat, C.; Merlin, S.; Pujadas, V.

    1997-01-01

    A worldwide increasing interest is presently observed for the waste vitrification whether they are radioactive or hazardous. Vitrification confines the waste in a stable and inert material and reduces significantly the waste volume which has a major effect on the disposal cost. The waste vitrification has been primarily applied for the treatment of high level radioactive waste from spent fuels reprocessing. In France, the CEA had a significant contribution in that field by developing in the 60's a technology based on metallic crucible heated by induction. The CEA continued to be actively engaged in an R and D effort and, since the 80's, is developing an advanced technology based on a cold crucible heated by induction. This technology particularly well fits with the requirements associated with LAW/Man waste treatment. Laboratory as well as preliminary full scale tests have been conducted with encouraging results to investigate the feasibility of direct ion exchange resins vitrification in a cold crucible. KEPRI investigated, In the past years, the different high temperature technologies which were available on the market and able to treat the low- and medium-level active waste produced by the NPP. The most promising technologies identified as a result of the studies were the cold crucible melter (CCM) for the conditioning of the evaporator concentrate, the ions exchange resins and the solid combustible waste and the plasma torch for the remaining solid waste such as filters

  9. Equipment experience in a radioactive LFCM [liquid-fed ceramic melter] vitrification facility

    International Nuclear Information System (INIS)

    Holton, L.K. Jr.; Dierks, R.D.; Sevigny, G.J.; Goles, R.W.; Surma, J.E.; Thomas, N.M.

    1986-11-01

    Since October 1984, the Pacific Northwest Laboratory (PNL) has operated a pilot-scale radioactive liquid-fed ceramic melter (RLFCM) vitrification process in shielded manipulator hot cells. This vitrification facility is being operated for the Department of Energy (DOE) to remotely test vitrification equipment components in a radioactive environment and to develop design and operation data that can be applied to production-scale projects. This paper summarizes equipment and process experience obtained from the operations of equipment systems for waste feeding, waste vitrification, canister filling, canister handling, and vitrification off-gas treatment

  10. 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

  11. 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.)

  12. Buried Waste Integrated Demonstration FY-93 Deployment Plan

    International Nuclear Information System (INIS)

    Bonnenberg, R.W.; Heard, R.E.; Milam, L.M.; Watson, L.R.

    1993-02-01

    The Buried Waste Integrated Demonstration (BWID) is a program funded by the US Department of Energy Office of Technology Development. BWID supports the applied research, development, demonstration, and evaluation of a suite of advanced technologies that together form a comprehensive remediation system for the effective and efficient remediation of buried waste. The fiscal year 1993 effort will deploy seven major field demonstrations at the Idaho National Engineering Laboratory's (INEL's) Radioactive Waste Management Complex Cold Test Pit. These major demonstrations are Remote Characterization System, Remote Excavation System, Overburden Removal, Waste Isolation, Contamination Control Unit, Rapid Monitoring Unit, and Fixation of Soil Surface Contamination. This document is the basic operational planning document for BWID deployment of the INEL field demonstrations. Additional sections deal briefly with four nonINEL field and laboratory demonstrations (Buried Waste Retrieval, Arc Melter Vitrification, Graphite DC Plasma Arc Melter, and Fixed Hearth Plasma Process) and with four INEL laboratory demonstrations (Electrostatic Curtain, Thermal Kinetics, Multiaxis Crane Control System, and Dig-Face Characterization)

  13. 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)

  14. 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

  15. 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.

  16. Characterization plan for the immobilized low-activity waste borehole

    International Nuclear Information System (INIS)

    Reidel, S.P.; Reynolds, K.D.

    1998-03-01

    The US Department of Energy's (DOE's) Hanford Site has the most diverse and largest amounts of radioactive tank waste in the US. High-level radioactive waste has been stored at Hanford in large underground tanks since 1944. Approximately 209,000 m 3 (54 Mgal) of waste are currently stored in 177 tanks. Vitrification and onsite disposal of low activity tank waste (LAW) are embodied in the strategy described in the Tri-Party Agreement. The tank waste is to be retrieved, separated into low- and high-level fractions, and then immobilized by private vendors. The DOE will receive the vitrified waste from private vendors and dispose of the low-activity fraction in the Hanford Site 200 East Area. The Immobilized Low-Activity Waste Disposal Complex (ILAWDC) is part of the disposal complex. This report is a plan to drill the first characterization borehole and collect data at the ILAWDC. This plan updates and revises the deep borehole portion of the characterization plan for the ILAWDC by Reidel and others (1995). It describes data collection activities for determining the physical and chemical properties of the vadose zone and the saturated zone at and in the immediate vicinity of the proposed ILAWDC. These properties then will be used to develop a conceptual geohydrologic model of the ILAWDC site in support of the Hanford ILAW Performance Assessment

  17. Vitrification Facility integrated system performance testing report

    International Nuclear Information System (INIS)

    Elliott, D.

    1997-01-01

    This report provides a summary of component and system performance testing associated with the Vitrification Facility (VF) following construction turnover. The VF at the West Valley Demonstration Project (WVDP) was designed to convert stored radioactive waste into a stable glass form for eventual disposal in a federal repository. Following an initial Functional and Checkout Testing of Systems (FACTS) Program and subsequent conversion of test stand equipment into the final VF, a testing program was executed to demonstrate successful performance of the components, subsystems, and systems that make up the vitrification process. Systems were started up and brought on line as construction was completed, until integrated system operation could be demonstrated to produce borosilicate glass using nonradioactive waste simulant. Integrated system testing and operation culminated with a successful Operational Readiness Review (ORR) and Department of Energy (DOE) approval to initiate vitrification of high-level waste (HLW) on June 19, 1996. Performance and integrated operational test runs conducted during the test program provided a means for critical examination, observation, and evaluation of the vitrification system. Test data taken for each Test Instruction Procedure (TIP) was used to evaluate component performance against system design and acceptance criteria, while test observations were used to correct, modify, or improve system operation. This process was critical in establishing operating conditions for the entire vitrification process

  18. 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

  19. Vitrification as an alternative to landfilling of tannery sewage sludge

    International Nuclear Information System (INIS)

    Celary, Piotr; Sobik-Szołtysek, Jolanta

    2014-01-01

    Highlights: • The possibility of vitrification of tannery sewage sludge was investigated. • Glass cullet was substituted with different wastes of mineral character. • Component ratio in the processed mixtures was optimized. • Environmental safety of the acquired vitrificates was verified. • An alternative management approach of usually landfilled waste was presented. - Abstract: Due to high content of heavy metals such as chromium, tannery sewage sludge is a material which is difficult to be biologically treated as it is in the case of organic waste. Consequently, a common practice in managing tannery sewage sludge is landfilling. This poses a potential threat to both soil and water environments and it additionally generates costs of construction of landfills that meet specific environment protection requirements. Vitrification of this kind of sewage sludge with the addition of mineral wastes can represent an alternative to landfilling. The aim of this study was to investigate the possibility of obtaining an environmentally safe product by means of vitrification of tannery sewage sludge from a flotation wastewater treatment process and chemical precipitation in order to address the upcoming issue of dealing with sewage sludge from the tannery industry which will be prohibited to be landfilled in Poland after 2016. The focus was set on determining mixtures of tannery sewage sludge with additives which would result in the lowest possible heavy metal leaching levels and highest hardness rating of the products obtained from their vitrification. The plasma vitrification process was carried out for mixtures with various amounts of additives depending on the type of sewage sludge used. Only the materials of waste character were used as additives. One finding of the study was an optimum content of mineral additives in vitrified mixture of 30% v/v waste molding sands with 20% v/v carbonate flotation waste from the zinc and lead industry for the formulations with

  20. Vitrification as an alternative to landfilling of tannery sewage sludge

    Energy Technology Data Exchange (ETDEWEB)

    Celary, Piotr, E-mail: pcelary@is.pcz.czest.pl; Sobik-Szołtysek, Jolanta, E-mail: jszoltysek@is.pcz.czest.pl

    2014-12-15

    Highlights: • The possibility of vitrification of tannery sewage sludge was investigated. • Glass cullet was substituted with different wastes of mineral character. • Component ratio in the processed mixtures was optimized. • Environmental safety of the acquired vitrificates was verified. • An alternative management approach of usually landfilled waste was presented. - Abstract: Due to high content of heavy metals such as chromium, tannery sewage sludge is a material which is difficult to be biologically treated as it is in the case of organic waste. Consequently, a common practice in managing tannery sewage sludge is landfilling. This poses a potential threat to both soil and water environments and it additionally generates costs of construction of landfills that meet specific environment protection requirements. Vitrification of this kind of sewage sludge with the addition of mineral wastes can represent an alternative to landfilling. The aim of this study was to investigate the possibility of obtaining an environmentally safe product by means of vitrification of tannery sewage sludge from a flotation wastewater treatment process and chemical precipitation in order to address the upcoming issue of dealing with sewage sludge from the tannery industry which will be prohibited to be landfilled in Poland after 2016. The focus was set on determining mixtures of tannery sewage sludge with additives which would result in the lowest possible heavy metal leaching levels and highest hardness rating of the products obtained from their vitrification. The plasma vitrification process was carried out for mixtures with various amounts of additives depending on the type of sewage sludge used. Only the materials of waste character were used as additives. One finding of the study was an optimum content of mineral additives in vitrified mixture of 30% v/v waste molding sands with 20% v/v carbonate flotation waste from the zinc and lead industry for the formulations with

  1. 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

  2. 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

  3. 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

  4. 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

  5. Laboratory optimization tests of technetium decontamination of Hanford Waste Treatment Plant low activity waste melter off-gas condensate simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M.L. [Savannah River Site (SRS), Aiken, SC (United States); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-11-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable simplified operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste.

  6. 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.

  7. 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

  8. 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

  9. ISV technology development plan for buried waste

    International Nuclear Information System (INIS)

    Nickelson, D.F.; Callow, R.A.; Luey, J.K.

    1992-07-01

    This report identifies the main technical issues facing the in situ vitrification (ISV) application to buried waste, and presents a plan showing the top-level schedule and projected resources needed to develop and demonstrate the technology for meeting Environmental Restoration Department (ERD) needs. The plan also proposes a model strategy for the technology transfer from the Department of Energy's Office of Technology Development (DOE-OTD) to the Office of Environmental Restoration (DOE-ER) as the technology proceeds from issues resolution (development) to demonstration and remedial readiness. Implementation of the plan would require $34,91 1K in total funding to be spread in the years FY-93 through FY-98. Of this amount, $10,183K is planned to be funded by DOE-OTD through the ISV Integrated Program. The remaining amount, $24,728K, is recommended to be split between the Department of Energy (DOE) Office of Technology Development ($6,670K) and DOE Office of Environmental Restoration ($18,058K)

  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. Feasibility testing of in situ vitrification of uranium-contaminated soils

    International Nuclear Information System (INIS)

    Ikuse, H.; Tsuchino, S.; Tasaka, H.; Timmerman, C.L.

    1989-01-01

    Process feasibility studies using in situ vitrification (ISV) were successfully performed on two different uranium-contaminated wastes. In situ vitrification is a thermal treatment process that converts contaminated soils into durable glass and crystalline form. Of the two different wastes, one waste was uranium mill tailings, while the other was uranium-contaminated soils which had high water contents. Analyses of the data from the two tests are presented

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. Laboratory Optimization Tests of Technetium Decontamination of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Melter Off-Gas Condensate Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-12-23

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste.

  17. 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

  18. Laboratory Optimization Tests of Decontamination of Cs, Sr, and Actinides from Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-01-06

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also substantially decrease the LAW vitrification mission duration and quantity of glass waste.

  19. ''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)

  20. 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.)

  1. 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

  2. Vitrification operational experiences and lessons learned at the WVDP

    International Nuclear Information System (INIS)

    Hamel, W.F. Jr.; Sheridan, M.J.; Valenti, P.J.

    1997-01-01

    The Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP) commenced full, high-level radioactive waste (HLW) processing activities in July 1996. The HLW consists of a blend of washed plutonium-uranium extraction (PUREX) sludge, neutralized thorium extraction (THOREX) waste, and cesium-loaded zeolite. The waste product is borosilicate glass contained in stainless steel canisters, sealed for eventual disposal in a federal repository. This paper discusses the WVDP vitrification process, focusing on operational experience and lessons learned during the first year of continuous, remote operation

  3. In situ vitrification program treatability investigation progress report

    International Nuclear Information System (INIS)

    Arrenholz, D.A.

    1991-02-01

    This document presents a summary of the efforts conducted under the in situ vitrification treatability study during the period from its initiation in FY-88 until FY-90. In situ vitrification is a thermal treatment process that uses electrical power to convert contaminated soils into a chemically inert and stable glass and crystalline product. Contaminants present in the soil are either incorporated into the product or are pyrolyzed during treatment. The treatability study being conducted at the Idaho National Engineering Laboratory by EG ampersand G Idaho is directed at examining the specific applicability of the in situ vitrification process to buried wastes contaminated with transuranic radionuclides and other contaminants found at the Subsurface Disposal Area of the Radioactive Waste Management Complex. This treatability study consists of a variety of tasks, including engineering tests, field tests, vitrified product evaluation, and analytical models of the in situ vitrification process. 6 refs., 4 figs., 3 tabs

  4. 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

  5. Testing of the West Valley Vitrification Facility transfer cart control system

    International Nuclear Information System (INIS)

    Halliwell, J.W.; Bradley, E.C.

    1995-01-01

    Oak Ridge National Laboratory (ORNL) has designed and tested the control system for the West Valley Demonstration Project Vitrification Facility transfer cart. The transfer cart will transfer canisters of vitrified high-level waste remotely within the Vitrification Facility. The control system operates the cart under battery power by wireless control. The equipment includes cart-mounted control electronics, battery charger, control pendants, engineer's console, and facility antennas. Testing was performed in several phases of development: (1) prototype equipment was built and tested during design, (2) board-level testing was then performed at ORNL during fabrication, and (3) system-level testing was then performed by ORNL at the fabrication subcontractor's facility for the completed cart system. These tests verified (1) the performance of the cart relative to design requirements and (2) operation of various built-in cart features. The final phase of testing is planned to be conducted during installation at the West Valley Vitrification Facility

  6. Office of River Protection Advanced Low-Activity Waste Glass Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, David K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kim, Dong-Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Piepel, Gregory F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-11-01

    The U.S. Department of Energy Office of River Protection (ORP) has initiated and leads an integrated Advanced Waste Glass (AWG) program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product performance requirements. The integrated ORP program is focused on providing a technical, science-based foundation for making key decisions regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities in the context of an optimized River Protection Project (RPP) flowsheet. The fundamental data stemming from this program will support development of advanced glass formulations, key product performance and process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste vitrification facilities. These activities will be conducted with the objective of improving the overall RPP mission by enhancing flexibility and reducing cost and schedule. The purpose of this advanced LAW glass research and development plan is to identify the near-term, mid-term, and longer-term research and development activities required to develop and validate advanced LAW glasses, property-composition models and their uncertainties, and an advanced glass algorithm to support WTP facility operations, including both Direct Feed LAW and full pretreatment flowsheets. Data are needed to develop, validate, and implement 1) new glass property-composition models and 2) a new glass formulation algorithm. Hence, this plan integrates specific studies associated with increasing the Na2O and SO3/halide concentrations in glass, because these components will ultimately dictate waste loadings for LAW vitrification. Of equal importance is the development of an efficient and economic strategy for 99Tc management. Specific and detailed studies are being implemented to understand the fate of Tc throughout

  7. 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

  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. 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 ...

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. 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

  16. Vitrification in the presence of salts

    International Nuclear Information System (INIS)

    Marra, J.C.; Andrews, M.K.; Schumacher, R.F.

    1994-01-01

    Glass is an advantageous material for the immobilization of nuclear wastes because of the simplicity of processing and its unique ability to accept a wide variety of waste elements into its network structure. Unfortunately, some anionic species which are present in the nuclear waste streams have only limited solubility in oxide glasses. This can result in either vitrification concerns or it can affect the integrity, of the final vitrified waste form. The presence of immiscible salts can also corrode metals and refractories in the vitrification unit as well as degrade components in the off-gas system. The presence of a molten salt layer on the melt may alter the batch melting rate and increase operational safety concerns. These safety concerns relate to the interaction of the molten salt and the melter cooling fluids. Some preliminary data from ongoing experimental efforts examining the solubility of molten salts in glasses and the interaction of salts with melter component materials is included

  17. 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

  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. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Serne, R. Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pierce, Eric M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chung, Chul-Woo [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Swanberg, David J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-05-31

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

  1. Functional description of the West Valley Demonstration Project Vitrification Facility

    International Nuclear Information System (INIS)

    Borisch, R.R.; McMahon, C.L.

    1990-07-01

    The primary objective of the West Valley Demonstration Project (WVDP) is the solidification of approximately 2.1 million liters (560,000 gallons) of high-level radioactive waste (HLW) which resulted from the operation of a nuclear fuel reprocessing plant. Since the original plant was not built to accommodate the processing of waste beyond storage in underground tanks, HLW solidification by vitrification presented numerous engineering challenges. Existing facilities required redesign and conversion to meet their new purpose. Vitrification technology and systems needed to be created and then tested. Equipment modifications, identified from cold test results, were incorporated into the final equipment configuration to be used for radioactive (hot) operations. Cold operations have defined the correct sequence and optimal functioning of the equipment to be used for vitrification and have verified the process by which waste will be solidified into borosilicate glass

  2. Stabilization of contaminated soils by in situ vitrification

    International Nuclear Information System (INIS)

    Timmerman, C.L.

    1984-01-01

    In Situ Vitrification is an emerging technology developed by Pacific Northwest Laboratory for potential in-place immobilization of radioactive wastes. The contaminated soil is stabilized and converted to an inert glass form. This conversion is accomplished by inserting electrodes in the soil and establishing an electric current between the electrodes. The electrical energy causes a joule heating effect that melts the soil during processing. Any contaminants released from the melt are collected and routed to an off-gas treatment system. A stable and durable glass block is produced which chemically and physically encapsulates any residual waste components. In situ vitrification has been developed for the potential application to radioactive wastes, specifically, contaminated soil sites; however, it could possibly be applied to hazardous chemical and buried munitions waste sites. The technology has been developed and demonstrated to date through a series of 21 engineering-scale tests [producing 50 to 1000 kg (100 to 2000 lb) blocks] and seven pilot-scale tests [producing 9000 kg (20,000 lb) blocks], the most recent of which illustrated treatment of actual radioactively contaminated soil. Testing with some organic materials has shown relatively complete thermal destruction and incineration. Further experiments have documented the insensitivity of in situ vitrification to soil characteristics such as fusion temperature, specific heat, thermal conductivity, electrical resistivity, and moisture content. Soil inclusions such as metals, cements, ceramics, and combustibles normally present only minor process limitations. Costs for hazardous waste applications are estimated to be less than $175/m 3 ($5.00/ft 3 ) of material vitrified. For many applications, in situ vitrification can provide a cost-effective alternative to other disposal options. 13 references, 4 figures, 1 table

  3. 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.

  4. Advanced High-Level Waste Glass Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, David K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-01

    The U.S. Department of Energy Office of River Protection (ORP) has implemented an integrated program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. The integrated ORP program is focused on providing a technical, science-based foundation from which key decisions can be made regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities. The fundamental data stemming from this program will support development of advanced glass formulations, key process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste (HLW) vitrification facilities with an appreciation toward reducing overall mission life. The purpose of this advanced HLW glass research and development plan is to identify the near-, mid-, and longer-term research and development activities required to develop and validate advanced HLW glasses and their associated models to support facility operations at WTP, including both direct feed and full pretreatment flowsheets. This plan also integrates technical support of facility operations and waste qualification activities to show the interdependence of these activities with the advanced waste glass (AWG) program to support the full WTP mission. Figure ES-1 shows these key ORP programmatic activities and their interfaces with both WTP facility operations and qualification needs. The plan is a living document that will be updated to reflect key advancements and mission strategy changes. The research outlined here is motivated by the potential for substantial economic benefits (e.g., significant increases in waste throughput and reductions in glass volumes) that will be realized when advancements in glass formulation continue and models supporting facility operations are implemented. Developing and applying advanced

  5. 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

  6. In Situ Vitrification Engineering-Scale Test ES-INEL-4, ES-INEL-5, ES-INEL-6, and ES-INEL-7 Test Plan

    International Nuclear Information System (INIS)

    Weidner, J.R.; Stoots, P.R.

    1990-10-01

    The In Situ Vitrification Engineering-Scale Tests ES-4, ES-5, ES-6, and ES-7 Product Characterization Test Plan describes the methods and procedures to be used or the physical and chemical characterization of the solid product(s) resulting from the Idaho National Engineering Laboratory engineering scale in situ vitrification tests ES-4, ES-5, ES-6, and ES-7. The goals of this Test Plan are to insure that the product characterization results are sufficient to meet the data needs of the In Situ Vitrification Program and are technically and legally defensible. Important issues addressed by the test plan include sampling and analysis strategy, sampling procedures, laboratory analysis, sample control and document management, equipment, data reporting and validation, quality assurance, specific routine procedures to assess data representativeness, safety and training program, and data management. 9 refs., 1 fig., 3 tabs

  7. Vitrification as an alternative to landfilling of tannery sewage sludge.

    Science.gov (United States)

    Celary, Piotr; Sobik-Szołtysek, Jolanta

    2014-12-01

    Due to high content of heavy metals such as chromium, tannery sewage sludge is a material which is difficult to be biologically treated as it is in the case of organic waste. Consequently, a common practice in managing tannery sewage sludge is landfilling. This poses a potential threat to both soil and water environments and it additionally generates costs of construction of landfills that meet specific environment protection requirements. Vitrification of this kind of sewage sludge with the addition of mineral wastes can represent an alternative to landfilling. The aim of this study was to investigate the possibility of obtaining an environmentally safe product by means of vitrification of tannery sewage sludge from a flotation wastewater treatment process and chemical precipitation in order to address the upcoming issue of dealing with sewage sludge from the tannery industry which will be prohibited to be landfilled in Poland after 2016. The focus was set on determining mixtures of tannery sewage sludge with additives which would result in the lowest possible heavy metal leaching levels and highest hardness rating of the products obtained from their vitrification. The plasma vitrification process was carried out for mixtures with various amounts of additives depending on the type of sewage sludge used. Only the materials of waste character were used as additives. One finding of the study was an optimum content of mineral additives in vitrified mixture of 30% v/v waste molding sands with 20% v/v carbonate flotation waste from the zinc and lead industry for the formulations with flotation sewage sludge, and 45% v/v and 5% v/v, respectively, for precipitation sewage sludge. These combinations allowed for obtaining products with negligible heavy metal leaching levels and hardness similar to commercial glass, which suggests they could be potentially used as construction aggregate substitutes. Incineration of sewage sludge before the vitrification process lead to

  8. 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

  9. 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

  10. Process arrangement options for Defense waste immobilization

    International Nuclear Information System (INIS)

    1980-02-01

    Current plans are to immobilize the SRP high-level liquid wastes in a high integrity form. Borosilicate glass was selected in 1977 as the reference waste form and a mjaor effort is currently underway to develop the required technology. A large new facility, referred to as the Defense Waste Processing Facility (DWPF) is being designed to carry out this mission, with project authorization targeted for 1982 and plant startup in 1989. However, a number of other process arrangements or manufacturing strategies, including staging the major elements of the project or using existing SRP facilities for some functions, have been suggested in lieu of building the reference DWPF. This study assesses these various options and compares them on a technical and cost basis with the DWPF. Eleven different manufacturing options for SRP defense waste solidification were examined in detail. These cases are: (1) vitrification of acid waste at current generation rate; (2) vitrification of current rate acid waste and caustic sludge; (3 and 4) vitrification of the sludge portion of neutralized waste; (5) decontamination of salt cake and storage of concentrated cesium and strontium for later immobilization; (6) processing waste in a facility with lower capacity than the DWPF; (7) processing waste in a combination of existing and new facilities; (8) waste immobilization in H Canyon; (9) vitrification of both sludge and salt; (10) DWPF with onsite storage; (11) deferred authorization of DWPF

  11. Vitrification of ion-exchange (IEX) resins: Advantages and technical challenges

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Peeler, D.K.; Cicero, C.A.

    1995-01-01

    Technologies are being developed by the US Department of Energy's (DOE) Savannah River Site (SRS) in conjunction with the Electric Power Research Institute (EPRI) and the commercial sector to convert low-level radioactive ion exchange (IEX) resin wastes from the nuclear utilities to solid stabilized waste forms for permanent disposal. One of the alternative waste stabilization technologies is vitrification of the resin into glass. Wastes can be vitrified at elevated temperatures by thermal treatment. One alternative thermal treatment is conventional Joule heated melting. Vitrification of wastes into glass is an attractive option because it atomistically bonds both hazardous and radioactive species in the glass structure, and volume reduces the wastes by 70-80%. The large volume reductions allow for large associated savings in disposal and/or long term storage costs

  12. 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

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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

  18. 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

  19. 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.)

  20. 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

  1. 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

  2. Feed Variability and Bulk Vitrification Glass Performance Assessment

    International Nuclear Information System (INIS)

    Mahoney, Lenna A.; Vienna, John D.

    2005-01-01

    The supplemental treatment (ST) bulk vitrification process will obtain its feed, consisting of low-activity waste (LAW), from more than one source. One purpose of this letter report is to describe the compositional variability of the feed to ST. The other is to support the M-62-08 decision by providing a preliminary assessment of the effectiveness of bulk vitrification (BV), the process that has been selected to perform supplemental treatment, in handling the ST feed envelope. Roughly nine-tenths of the ST LAW feed will come from the Waste Treatment Plant (WTP) pretreatment. This processed waste is expected to combine (1) a portion of the same LAW feed sent to the WTP melters and (2) a dilute stream that is the product of the condensate from the submerged-bed scrubber (SBS) and the drainage from the electrostatic precipitator (WESP), both of which are part of the LAW off-gas system. The manner in which the off-gas-product stream is concentrated to reduce its volume, and the way in which the excess LAW and off-gas product streams are combined, are part of the interface between WTP and ST and have not been determined. This letter report considers only one possible arrangement, in which half of the total LAW is added to the off-gas product stream, giving an estimated ST feed stream from WTP. (Total LAW equals that portion of LAW sent to the WTP LAW vitrification plant (WTP LAW) plus the LAW not currently treatable in the LAW vitrification plant due to capacity limitations (excess))

  3. 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).

  4. 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)

  5. 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

  6. Computer simulation of the off gas treatment process for the KEPCO pilot vitrification plant

    International Nuclear Information System (INIS)

    Kim, Hey Suk; Maeng, Sung Jun; Lee, Myung Chan

    1999-01-01

    Vitrification technology for treatment of low and intermediate radioactive wastes can remarkably reduce waste volume to about one twentieth of the initial volume as they are collected and converted into a very stable form. Therefore, it can minimize environmental impact when the vitrified waste is disposed of. But an off gas treatment system is necessary to apply this technology because air pollutants and radioisotopes are generated like those of other conventional incinerators during thermal oxidation process at high temperature. KEPCO designed and installed a pilot scale vitrification plant to demonstrate the feasibility of the vitrification process and then to make a conceptual design for a commercial vitrification facility. The purpose of this study was to simulate the off gas treatment system(OGTS) in order optimize the operating conditions. Mass balance and temperature profile in the off gas treatment system were simulated for different combinations of combustible wastes by computer simulation code named OGTS code and removal efficiency of each process was also calculated with change of design parameters. The OGTS code saved efforts,time and capital because scale and configuration of the system could be easily changed. The simulation result of the pilot scale off gas process as well as pilot tests will be of great use in the future for a design of the commercial vitrification facility. (author)

  7. In situ vitrification program treatability investigation progress report

    International Nuclear Information System (INIS)

    Arrenholz, D.A.

    1990-12-01

    This document presents a summary of the efforts conducted under the in situ vitrification treatability study during the period from its initiation in FY-88 until FY-90. In situ vitrification is a thermal treatment process that uses electrical power to convert contaminated soils into a chemically inert and stable glass and crystalline product. Contaminants present in the soil are either incorporated into the product or are pyrolyzed during treatment. The treatability study being conducted at the Idaho National Engineering Laboratory by EG ampersand G Idaho is directed at examining the specific applicability of the in situ vitrification process to buried wastes contaminated with transuranic radionuclides and other contaminants found at the Subsurface Disposal Area of the Radioactive Waste Management Complex. This treatability study consists of a variety of tasks, including engineering tests, field tests, vitrified product evaluation, and analytical models of the ISV process. The data collected in the course of these efforts will address the nine criteria set forth in the Comprehensive Environmental Response, Compensation, and Liability Act, which will be used to identify and select specific technologies to be used in the remediation of the buried wastes at the Subsurface Disposal Area. 6 refs., 4 figs., 3 tabs

  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. 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

  11. 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

  12. 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

  13. Vitrification testing of soil fines from contaminated Hanford 100 Area and 300 Area soils

    International Nuclear Information System (INIS)

    Ludowise, J.D.

    1994-01-01

    The suitability of Hanford soil for vitrification is well known and has been demonstrated extensively in other work. The tests reported here were carried out to confirm the applicability of vitrification to the soil fines (a subset of the Hanford soil potentially different in composition from the bulk soil) and to provide data on the performance of actual, vitrified soil fines. It was determined that the soil fines were generally similar in composition to the bulk Hanford soil, although the fraction 2 O. The vitrified waste (plus additives) occupies only 60% of the volume of the initial untreated waste. Leach testing has shown the glasses made from the soil fines to be very durable relative to natural and man-made glasses and has demonstrated the ability of the vitrified waste to greatly reduce the release of radionuclides to the environment. Viscosity and electrical conductivity measurements indicate that the soil fines will be readily processable, although with levels of additives slightly greater than used in the radioactive melts. These tests demonstrate the applicability of vitrification to the contaminated soil fines and the exceptional performance of the waste form resulting from the vitrification of contaminated Hanford soils

  14. Test plan: Effects of phase separation on waste loading for high level waste glasses

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    2000-01-01

    As part of the Tanks Focus Area's (TFA) effort to increase waste loading for high-level waste (HLW) vitrification at various facilities in the Department of Energy (DOE) complex, the occurrence of phase separation in waste glasses spanning the Savannah River Site (SRS) and Idaho National Engineering and Environmental Laboratory (INEEL) composition ranges were studied during FY99. The type, extent, and impact of phase separation on glass durability for a series of HLW glasses, e.g., SRS-type and INEEL-type, were examined

  15. 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

  16. HWVP pilot-scale vitrification system campaign: LFCM-8 summary report

    International Nuclear Information System (INIS)

    Perez, J.M.; Whitney, L.D.; Buchmiller, W.C.; Daume, J.T.; Whyatt, G.A.

    1996-04-01

    The Hanford Waste Vitrification Plant (HWVP) is being designed to treat the high-level radiative waste (HLW) stored in underground storage tanks as an alkaline sludge. Tank waste will first be retrieved and pretreated to minimize solids requiring vitrification as HLW. The glass product resulting from HWVP operations will be stored onsite in stainless steel canisters until the HLW repository is available for final disposal. The first waste stream scheduled to be processed by the HWVP is the neutralized current acid waste (NCAW) stored in double-shell storage tanks. The Pacific Northwest Laboratory (PNL) is supporting Westinghouse Hanford Company (WHC) by providing research, development, and engineering expertise in defined areas. As a part of this support, pilot-scale testing is being conducted to support closure of HWVP design and development issues. Testing results will verify equipment design performance, establish acceptable and optimum process parameters, and support product qualification activities

  17. HWVP pilot-scale vitrification system campaign: LFCM-8 summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perez, J.M.; Whitney, L.D.; Buchmiller, W.C.; Daume, J.T.; Whyatt, G.A.

    1996-04-01

    The Hanford Waste Vitrification Plant (HWVP) is being designed to treat the high-level radiative waste (HLW) stored in underground storage tanks as an alkaline sludge. Tank waste will first be retrieved and pretreated to minimize solids requiring vitrification as HLW. The glass product resulting from HWVP operations will be stored onsite in stainless steel canisters until the HLW repository is available for final disposal. The first waste stream scheduled to be processed by the HWVP is the neutralized current acid waste (NCAW) stored in double-shell storage tanks. The Pacific Northwest Laboratory (PNL) is supporting Westinghouse Hanford Company (WHC) by providing research, development, and engineering expertise in defined areas. As a part of this support, pilot-scale testing is being conducted to support closure of HWVP design and development issues. Testing results will verify equipment design performance, establish acceptable and optimum process parameters, and support product qualification activities.

  18. Development of analytical cell support for vitrification at the West Valley Demonstration Project. Topical report

    Energy Technology Data Exchange (ETDEWEB)

    Barber, F.H.; Borek, T.T.; Christopher, J.Z. [and others

    1997-12-01

    Analytical and Process Chemistry (A&PC) support is essential to the high-level waste vitrification campaign at the West Valley Demonstration Project (WVDP). A&PC characterizes the waste, providing information necessary to formulate the recipe for the target radioactive glass product. High-level waste (HLW) samples are prepared and analyzed in the analytical cells (ACs) and Sample Storage Cell (SSC) on the third floor of the main plant. The high levels of radioactivity in the samples require handling them in the shielded cells with remote manipulators. The analytical hot cells and third floor laboratories were refurbished to ensure optimal uninterrupted operation during the vitrification campaign. New and modified instrumentation, tools, sample preparation and analysis techniques, and equipment and training were required for A&PC to support vitrification. Analytical Cell Mockup Units (ACMUs) were designed to facilitate method development, scientist and technician training, and planning for analytical process flow. The ACMUs were fabricated and installed to simulate the analytical cell environment and dimensions. New techniques, equipment, and tools could be evaluated m in the ACMUs without the consequences of generating or handling radioactive waste. Tools were fabricated, handling and disposal of wastes was addressed, and spatial arrangements for equipment were refined. As a result of the work at the ACMUs the remote preparation and analysis methods and the equipment and tools were ready for installation into the ACs and SSC m in July 1995. Before use m in the hot cells, all remote methods had been validated and four to eight technicians were trained on each. Fine tuning of the procedures has been ongoing at the ACs based on input from A&PC technicians. Working at the ACs presents greater challenges than had development at the ACMUs. The ACMU work and further refinements m in the ACs have resulted m in a reduction m in analysis turnaround time (TAT).

  19. The R7/T7 vitrification at La Hague: 10 years of operation

    International Nuclear Information System (INIS)

    Masson, H.; Desvaux, J.L.; Pluche, E.; Jouan, A.

    2001-01-01

    Vitrification of high level wastes from reprocessing of spent nuclear fuels has been carried out at La Hague on an industrial scale for ten years. This paper presents an historical overview of the facilities, and describes the facilities and their operations, startup performance, facility upgrading that has been done, and process control functions. The paper concludes that the technology for vitrification of high level wastes is mature and has been mastered. (author)

  20. 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.

  1. Radiation exposure control by estimation of multiplication factors for online remote radiation monitoring systems at vitrification plant

    International Nuclear Information System (INIS)

    Deokar, U.V.; Kulkarni, V.V.; Khot, A.R.; Mathew, P.; Kamlesh; Purohit, R.G.; Sarkar, P.K.

    2012-01-01

    Vitrification Plant is commissioned for vitrification of high level liquid waste (HLW) generated in nuclear fuel cycle operations by using Joule Heated Ceramic Melter first time in India. Exposure control is a major concern in operating plant. Therefore in addition to installed monitors, we have developed online remote radiation monitoring system to minimize number of entries in amber areas and to reduce the exposure to the surveyor and operator. This also helped in volume reduction of secondary waste. The reliability and accuracy of the online monitoring system is confirmed with actual measurements and by theoretical shielding calculations. The multiplication factors were estimated for remote on line monitoring of Melter Off Gas (MOG) filter, Hood filter, three exhaust filter banks, and over-pack monitoring. This paper summarizes - how the online remote monitoring system helped in saving of 128.52 person-mSv collective dose (14.28% of budgeted dose). The system also helped in the reduction of 2.6 m 3 of Cat-I waste. Our online remote monitoring system has helped the plant management to plan in advance for replacement of these filters, which resulted in considerable saving in collective dose and secondary waste

  2. 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)

  3. 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)

  4. 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)

  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. 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

  7. A COMPREHENSIVE TECHNICAL REVIEW OF THE DEMONSTRATION BULK VITRIFICATION SYSTEM

    International Nuclear Information System (INIS)

    SCHAUS, P.S.

    2006-01-01

    In May 2006, CH2M Hill Hanford Group, Inc. chartered an Expert Review Panel (ERP) to review the current status of the Demonstration Bulk Vitrification System (DBVS). It is the consensus of the ERP that bulk vitrification is a technology that requires further development and evaluation to determine its potential for meeting the Hanford waste stabilization mission. No fatal flaws (issues that would jeopardize the overall DBVS mission that cannot be mitigated) were found, given the current state of the project. However, a number of technical issues were found that could significantly affect the project's ability to meet its overall mission as stated in the project ''Justification of Mission Need'' document, if not satisfactorily resolved. The ERP recognizes that the project has changed from an accelerated schedule demonstration project to a formally chartered project that must be in full compliance with DOE 413.3 requirements. The perspective of the ERP presented herein, is measured against the formally chartered project as stated in the approved Justification of Mission Need document. A justification of Mission Need document was approved in July 2006 which defined the objectives for the DBVS Project. In this document, DOE concluded that bulk vitrification is a viable technology that requires additional development to determine its potential applicability to treatment of a portion of the Hanford low activity waste. The DBVS mission need statement now includes the following primary objectives: (1) process approximately 190,000 gallons of Tank S-109 waste into fifty 100 metric ton boxes of vitrified product; (2) store and dispose of these boxes at Hanford's Integrated Disposal Facility (IDF); (3) evaluate the waste form characteristics; (4) gather pilot plant operability data, and (5) develop the overall life cycle system performance of bulk vitrification and produce a comparison of the bulk vitrification process to building a second LAW Immobilization facility or other

  8. 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

  9. 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.

  10. High-level waste melter alternatives assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, R.B.

    1995-02-01

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

  11. High-level waste melter alternatives assessment report

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-02-01

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

  12. 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

  13. Bench scale experiments for the remediation of Hanford Waste Treatment Plant low activity waste melter off-gas condensate

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M.L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Poirier, Michael [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-08-11

    The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The plan for disposition of this stream during baseline operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. This stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the LAW vitrification mission duration and quantity of glass waste. The concept being tested here involves removing the 99Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed elsewhere.

  14. Evaluation of cold testing for Tokai Vitrification Facility

    International Nuclear Information System (INIS)

    Yoshioka, Masahiro; Inada, Eiichi

    1994-01-01

    The cold testing of the Tokai Vitrification Facility (TVF) was completed at the end of March, 1994 through the tests of nearly two years since May in 1992. The cold testing was carried out in order to evaluate the process equipment, product quality control, remote maintenance capability. The test results shown that TVF has enough performance with safety to treat the liquid waste in each process, and to control the product quality. For the remote maintenance of process equipment in the vitrification cell, the remote maintenance capability was confirmed for all remote equipment in the cell. The improvements were taken for some equipment with problem from the point of the operability and maintenance. It was confirmed by these test results that the TVF can go forward to the hot test operation using actual waste. (author)

  15. 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)

  16. Innovative vitrification for soil remediation

    International Nuclear Information System (INIS)

    Jetta, N.W.; Patten, J.S.; Hart, J.G.

    1995-01-01

    The objective of this DOE demonstration program is to validate the performance and operation of the Vortec Cyclone Melting System (CMS trademark) for the processing of LLW contaminated soils found at DOE sites. This DOE vitrification demonstration project has successfully progressed through the first two phases. Phase 1 consisted of pilot scale testing with surrogate wastes and the conceptual design of a process plant operating at a generic DOE site. The objective of Phase 2, which is scheduled to be completed the end of FY 95, is to develop a definitive process plant design for the treatment of wastes at a specific DOE facility. During Phase 2, a site specific design was developed for the processing of LLW soils and muds containing TSCA organics and RCRA metal contaminants. Phase 3 will consist of a full scale demonstration at the DOE gaseous diffusion plant located in Paducah, KY. Several DOE sites were evaluated for potential application of the technology. Paducah was selected for the demonstration program because of their urgent waste remediation needs as well as their strong management and cost sharing financial support for the project. During Phase 2, the basic nitrification process design was modified to meet the specific needs of the new waste streams available at Paducah. The system design developed for Paducah has significantly enhanced the processing capabilities of the Vortec vitrification process. The overall system design now includes the capability to shred entire drums and drum packs containing mud, concrete, plastics and PCB's as well as bulk waste materials. This enhanced processing capability will substantially expand the total DOE waste remediation applications of the technology

  17. LFCM [liquid-fed ceramic melter] vitrification technology: Quarterly progress report, January--March 1987

    International Nuclear Information System (INIS)

    Brouns, R. A.; Allen, C. R.; Powell, J. A.

    1988-05-01

    This report is compiled by the Nuclear Waste Treatment Program and the Hanford Waste Vitrification Program at Pacific Northwest Laboratory to describe the progress in developing, testing, applying and documenting liquid-fed ceramic melter vitrification technology. Progress in the following technical subject areas during the second quarter of FY 1987 is discussed: melting process chemistry and glass development, feed preparation and transfer systems, melter systems, canister filling and handling systems, and process/product modeling. 23 refs., 14 figs., 10 tabs

  18. 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

  19. 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

  20. 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

  1. High-level waste processing and disposal

    International Nuclear Information System (INIS)

    Crandall, J.L.; Krause, H.; Sombret, C.; Uematsu, K.

    1984-11-01

    Without reprocessing, spent LWR fuel itself is generally considered an acceptable waste form. With reprocessing, borosilicate glass canisters, have now gained general acceptance for waste immobilization. The current first choice for disposal is emplacement in an engineered structure in a mined cavern at a depth of 500-1000 meters. A variety of rock types are being investigated including basalt, clay, granite, salt, shale, and volcanic tuff. This paper gives specific coverage to the national high level waste disposal plans for France, the Federal Republic of Germany, Japan and the United States. The French nuclear program assumes prompt reprocessing of its spent fuels, and France has already constructed the AVM. Two larger borosilicate glass plants are planned for a new French reprocessing plant at La Hague. France plans to hold the glass canisters in near-surface storage for a forty to sixty year cooling period and then to place them into a mined repository. The FRG and Japan also plan reprocessing for their LWR fuels. Both are currently having some fuel reprocessed by France, but both are also planning reprocessing plants which will include waste vitrification facilities. West Germany is now constructing the PAMELA Plant at Mol, Belgium to vitrify high level reprocessing wastes at the shutdown Eurochemic Plant. Japan is now operating a vitrification mockup test facility and plans a pilot plant facility at the Tokai reprocessing plant by 1990. Both countries have active geologic repository programs. The United State program assumes little LWR fuel reprocessing and is thus primarily aimed at direct disposal of spent fuel into mined repositories. However, the US have two borosilicate glass plants under construction to vitrify existing reprocessing wastes

  2. Vitrification of spent mordenite molecular sieves

    International Nuclear Information System (INIS)

    Sathi Sasidharan, N.; Deshingkar, D.S.; Jain, Savita; Singh, I.J.; Wattal, P.K.

    2002-11-01

    Vitrification of cesium loaded inorganic ion exchangers (mordenite type molecular sieves/zeolite AR-1) was studied empolying borosilicate glass systems. Direct vitrification of aluminosilicates is rather difficult mainly on account of volatility of cesium at processing temperatures of 1100 degC-1300 degC. In the borosilicate glass system, oxides of lead, sodium and zinc along with boric oxide were employed as major glass formers. Homogeneous glass matrix was obtained incorporating simulated composition of mordenite along with oxides of sodium, lead and boron at the processing temperature of 950 degC. The waste oxide loading up to 50% on dry weight basis was incorporated in this glass formulation. Partial replacement of PbO by TeO 2 , Bi 2 O 3 and CaF 2 resulted in lowering of the processing temperature and also increasing homogeneity of matrix. Based on these results, a glass matrix was prepared with actual cesium AR-1 molecular sieves with processing temperature limited to 925 degC. Powdered samples of glass matrix were subjected to leaching as per ASTM-1285 Product Consistency Test in high purity water at 90 degC for 28 days. The normalised cesium leach rate of this glass was found to be 3.92 x 10 -6 g/cm 2 /day, which is comparable to sodium borosilicate glass matrices currently in use for immobilisation of high level waste. The molecular sieves are also amenable to immobilization in cement matrix. As expected, there is substantial volume reduction by factor 3 in vitrification compared to their immobilization in cementious matrices. Also the quantity of cesium leached from vitrified product was nearly 10,000 times lower compared to cement based matrix. Vitrification of mordenite molecular sieves would lead to high capacity utilisation of zeolite AR-1 for the treatment of low and intennediate levelliquid effluents. (author)

  3. Overview of the West Valley Vitrification Facility transfer cart control system

    International Nuclear Information System (INIS)

    Bradley, E.C.; Rupple, F.R.

    1993-01-01

    Oak Ridge National Laboratory (ORNL) has designed the control system for the West Valley Demonstration Project Vitrification Facility transfer cart. The transfer cart will transfer canisters of vitrified high-level waste remotely within the Vitrification Facility. The control system will operate the cart under battery power by wireless control. The equipment includes cart mounted control electronics, battery charger, control pendants, engineer's console, and facility antennas

  4. India gets set at Tarapur [vitrification plant

    International Nuclear Information System (INIS)

    Cruickshank, Andrew.

    1987-01-01

    A vitrification plant has been built and commissioned at Tarapur to immobilise high level radioactive waste arising from the reprocessing plant. The plant employs a semi-continuous pot-glass process, involving calcination followed by melting in the processing vessel and subsequent casting of the glass in a storage container. Prior to disposal the waste is stored in an air-cooled vault with a convective air-circulation system. (author)

  5. A pilot plant demonstration of the vitrification of radioactive solutions using microwave power

    International Nuclear Information System (INIS)

    Morrell, M.S.; Hardwick, W.H.; Murphy, V.; Wace, P.F.

    1986-01-01

    A process has been developed that exploits the characteristics of microwave heating for the vitrification of high-level radioactive liquid waste. This process, microwave vitrification, has been successfully operated at pilot plant scale in an active cell using simulated liquid waste containing several curies of radioactivity. Excellent decontamination factors have been achieved for both volatiles and nonvolatiles with an average ruthenium decontamination factor of 490 and a gross alpha emitter decontamination factor of 100,000. Almost all the radioactivity is incorporated in a glass block

  6. 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

  7. 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...

  8. 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.

  9. 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

  10. Online remote radiological monitoring during operation of Advance Vitrification System (AVS), Tarapur

    International Nuclear Information System (INIS)

    Deokar, U.V.; Kulkarni, V.V.; Mathew, P.; Khot, A.R.; Singh, K.K.; Kamlesh; Deshpande, M.D.; Kulkarni, Y.

    2010-01-01

    Advanced Vitrification System (AVS) is commissioned for vitrification of high level waste (HLW) by using Joule heated ceramic melter first time in India. The HLW is generated in fuel reprocessing plant. For radiological surveillance of plant, Health Physics Unit (HPU) had installed 37 Area Gamma Monitors (AGM), 7 Continuous Air Monitors (CAM) and all types of personal contamination monitors. Exposure control is a major concern in operating plant. Therefore in addition to installed monitors, we have developed online remote radiation monitoring system to minimize exposures to the surveyor and operator. This also helped in volume reduction of secondary waste. The reliability and accuracy of the online monitoring system is confirmed by calibrating the system by comparing TLD and DRD readings and by theoretical analysis. In addition some modifications were carried in HP instruments to make them user friendly. This paper summarizes different kinds of remote radiological monitoring systems installed for online monitoring of Melter off Gas (MOG) filter, Hood filter, three exhaust filter banks, annulus air sampling and over pack monitoring in AVS. Our online remote monitoring system has helped the plant management to plan in advance for replacement of these filters, which resulted in considerable saving of collective dose. (author)

  11. 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.

  12. 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

  13. The design and construction of the windscale vitrification plant and vitrified product store

    International Nuclear Information System (INIS)

    Heafield, W.; Woodall, A.; Elsden, A.D.

    1987-01-01

    The paper describes the background of High Level Waste storage and vitrification development in the UK and its application to Reprocessing Operations at Sellafield. The main stages in the vitrification process and associated maintenance facilities are described together with the layout of the Windscale Vitrification Plant (WVP) and associated Vitrified Product Store (VPS). The design and construction techniques employed for example, the use of Computer Aided Design and the effect of automatic pipe bending/orbital welding and the use of precast units for cell construction, are discussed and current construction progress is highlighted. The vitrification process uses complex mechanical plant operating in high temperature and radiation fields. An extensive engineering and process development programme has been carried out. A full scale inactive facility (FSIF) has been constructed and the objectives and results from the operation of FSIF are presented. In addition to engineering and process development, a comprehensive programme of glass technology development has been carried out to establish maximum waste incorporation levels, reaction kinetic and product properties of the candidate glass formulations

  14. Tank Waste Remediation System optimized processing strategy

    International Nuclear Information System (INIS)

    Slaathaug, E.J.; Boldt, A.L.; Boomer, K.D.; Galbraith, J.D.; Leach, C.E.; Waldo, T.L.

    1996-03-01

    This report provides an alternative strategy evolved from the current Hanford Site Tank Waste Remediation System (TWRS) programmatic baseline for accomplishing the treatment and disposal of the Hanford Site tank wastes. This optimized processing strategy performs the major elements of the TWRS Program, but modifies the deployment of selected treatment technologies to reduce the program cost. The present program for development of waste retrieval, pretreatment, and vitrification technologies continues, but the optimized processing strategy reuses a single facility to accomplish the separations/low-activity waste (LAW) vitrification and the high-level waste (HLW) vitrification processes sequentially, thereby eliminating the need for a separate HLW vitrification facility

  15. Solid waste handling

    International Nuclear Information System (INIS)

    Parazin, R.J.

    1995-01-01

    This study presents estimates of the solid radioactive waste quantities that will be generated in the Separations, Low-Level Waste Vitrification and High-Level Waste Vitrification facilities, collectively called the Tank Waste Remediation System Treatment Complex, over the life of these facilities. This study then considers previous estimates from other 200 Area generators and compares alternative methods of handling (segregation, packaging, assaying, shipping, etc.)

  16. 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

  17. INNOVATIVE FOSSIL FUEL FIRED VITRIFICATION TECHNOLOGY FOR SOIL REMEDIATION. SUMMARY REPORT

    International Nuclear Information System (INIS)

    J. Hnat; L.M. Bartone; M. Pineda

    2001-01-01

    This Summary Report summarizes the progress of Phases 3, 3A and 4 of a waste technology Demonstration Project sponsored under a DOE Environmental Management Research and Development Program and administered by the U.S. Department of Energy National Energy Technology Laboratory-Morgantown (DOE-NETL) for an ''Innovative Fossil Fuel Fired Vitrification Technology for Soil Remediation''. The Summary Reports for Phases 1 and 2 of the Program were previously submitted to DOE. The total scope of Phase 3 was to have included the design, construction and demonstration of Vortec's integrated waste pretreatment and vitrification process for the treatment of low level waste (LLW), TSCA/LLW and mixed low-level waste (MLLW). Due to funding limitations and delays in the project resulting from a law suit filed by an environmental activist and the extended time for DOE to complete an Environmental Assessment for the project, the scope of the project was reduced to completing the design, construction and testing of the front end of the process which consists of the Material Handling and Waste Conditioning (MH/C) Subsystem of the vitrification plant. Activities completed under Phases 3A and 4 addressed completion of the engineering, design and documentation of the Material Handling and Conditioning System such that final procurement of the remaining process assemblies can be completed and construction of a Limited Demonstration Project be initiated in the event DOE elects to proceed with the construction and demonstration testing of the MH/C Subsystem

  18. 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

  19. Development of analytical cell support for vitrification at the West Valley Demonstration Project. Topical report

    International Nuclear Information System (INIS)

    Barber, F.H.; Borek, T.T.; Christopher, J.Z.

    1997-12-01

    Analytical and Process Chemistry (A ampersand PC) support is essential to the high-level waste vitrification campaign at the West Valley Demonstration Project (WVDP). A ampersand PC characterizes the waste, providing information necessary to formulate the recipe for the target radioactive glass product. High-level waste (HLW) samples are prepared and analyzed in the analytical cells (ACs) and Sample Storage Cell (SSC) on the third floor of the main plant. The high levels of radioactivity in the samples require handling them in the shielded cells with remote manipulators. The analytical hot cells and third floor laboratories were refurbished to ensure optimal uninterrupted operation during the vitrification campaign. New and modified instrumentation, tools, sample preparation and analysis techniques, and equipment and training were required for A ampersand PC to support vitrification. Analytical Cell Mockup Units (ACMUs) were designed to facilitate method development, scientist and technician training, and planning for analytical process flow. The ACMUs were fabricated and installed to simulate the analytical cell environment and dimensions. New techniques, equipment, and tools could be evaluated m in the ACMUs without the consequences of generating or handling radioactive waste. Tools were fabricated, handling and disposal of wastes was addressed, and spatial arrangements for equipment were refined. As a result of the work at the ACMUs the remote preparation and analysis methods and the equipment and tools were ready for installation into the ACs and SSC m in July 1995. Before use m in the hot cells, all remote methods had been validated and four to eight technicians were trained on each. Fine tuning of the procedures has been ongoing at the ACs based on input from A ampersand PC technicians. Working at the ACs presents greater challenges than had development at the ACMUs. The ACMU work and further refinements m in the ACs have resulted m in a reduction m in

  20. Investigation of variable compositions on the removal of technetium from Hanford Waste Treatment Plant low activity waste melter off-gas condensate simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, John M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-03-29

    The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the offgas system. The plan for disposition of this stream during baseline operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. This stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the LAW vitrification mission duration and quantity of glass waste. The concept being tested here involves removing the 99Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed elsewhere.

  1. Quality Assurance Project Plan for the treatability study of in situ vitrification of Seepage Pit 1 in Waste Area Grouping 7 at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    1995-07-01

    This Quality Assurance Project Plan (QAPjP) establishes the quality assurance procedures and requirements to be implemented for the control of quality-related activities for Phase 3 of the Treatability Study (TS) of In Situ Vitrification (ISV) of Seepage Pit 1, ORNL Waste Area Grouping 7. This QAPjP supplements the Quality Assurance Plan for Oak Ridge National Laboratory Environmental Restoration Program by providing information specific to the ISV-TS. Phase 3 of the TS involves the actual ISV melt operations and posttest monitoring of Pit 1 and vicinity. Previously, Phase 1 activities were completed, which involved determining the boundaries of Pit 1, using driven rods and pipes and mapping the distribution of radioactivity using logging tools within the pipes. Phase 2 involved sampling the contents, both liquid and solids, in and around seepage Pit 1 to determine their chemical and radionuclide composition and the spatial distribution of these attributes. A separate QAPjP was developed for each phase of the project. A readiness review of the Phase 3 activities presented QAPjP will be conducted prior to initiating field activities, and an Operational Acceptance, Test (OAT) will also be conducted with no contamination involved. After, the OAT is complete, the ISV process will be restarted, and the melt will be allowed to increase with depth and incorporate the radionuclide contamination at the bottom of Pit 1. Upon completion of melt 1, the equipment will be shut down and mobilized to an adjacent location at which melt 2 will commence

  2. Tolerancing requirements for remote handling at the Hanford vitrification project

    International Nuclear Information System (INIS)

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

    1993-01-01

    The Hanford Waste Vitrification Plant is being designed by Fluor Daniel, Inc. with WasteChem Corporation as Fluor Daniel's major subcontractor specializing in vitrification and remote system technologies. United Engineers and Constructors/Catalytic (UE ampersand C) will construct the plant. Westinghouse Hanford Company (WHC) is the Project Integration manager, manager and as the plant operator provides technical direction to the Architect/Engineer team (A/E) and constructor on behalf of the Department of Energy - Richland Field Office. The A/E has developed, in cooperation with UE ampersand C, WHC and DOE, 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

  3. 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.

  4. 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

  5. Treatment of hazardous metals by in situ vitrification

    International Nuclear Information System (INIS)

    Koegler, S.S.; Buelt, J.L.

    1989-02-01

    Soils contaminated with hazardous metals are a significant problem to many Defense Program sites. Contaminated soils have ranked high in assessments of research and development needs conducted by the Hazardous Waste Remedial Action Program (HAZWRAP) in FY 1988 and FY 1989. In situ vitrification (ISV) is an innovative technology suitable for stabilizing soils contaminated with radionuclides and hazardous materials. Since ISV treats the material in place, it avoids costly and hazardous preprocessing exhumation of waste. In situ vitrification was originally developed for immobilizing radioactive (primarily transuranic) soil constituents. Tests indicate that it is highly useful also for treating other soil contaminants, including hazardous metals. The ISV process produces an environmentally acceptable, highly durable glasslike product. In addition, ISV includes an efficient off-gas treatment system that eliminates noxious gaseous emissions and generates minimal hazardous byproducts. This document reviews the Technical Basis of this technology. 5 refs., 7 figs., 2 tabs

  6. 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

  7. Commissioning Tests of the Ulchin LLW Vitrification Facility In Korea

    International Nuclear Information System (INIS)

    Kyung-Hwa, Yang; Sang-Woon, Shin; Chan-Kook, Moon

    2009-01-01

    Since 1994, Korea Hydro and Nuclear Power Co., Ltd. (KHNP) has, together with SGN in France and Hyundai ROTEM, investigated and developed a vitrification process using a Cold Crucible Induction Melter (CCIM) to treat low-and intermediate-level radioactive waste. A commercialization project was launched in 2002 as a governmental nuclear power technology development project. The installation of the first commercial plant, Ulchin Vitrification Facility (UVF), was completed in 2007 inside Ulchin nuclear power plants no. 5 and 6. Combustible dry active waste and low-level ion exchange resin will be treated in the UVF. The UVF has a waste feeding capacity of 20 kg/h and consists of waste pretreatment and feeding systems, a cold crucible induction melter (CCIM) system, an off-gas treatment system, a dust recycling system, as well as other systems. In order to assure that systems and equipments meet their design objectives and that the UVF complies with applicable regulations, equipment tests, system functional tests and inactive performance tests were conducted. Furthermore, a long-term inactive test was carried out for 202 hours to evaluate the overall performance and stability of the facility. During the test, about 1,700 kg of surrogate waste was vitrified and 302 kg of waste glass was poured into a glass mould. As the gaseous emission from the UVF was one of the key issues for the operational license and public acceptance, 25 hazardous gases and dusts were analyzed. The compressive strength of the waste glasses was also measured. Results showed that effluent concentrations of the off-gases and the quality of the waste glass met the regulatory limits with sufficient margins. Operation procedures of the UVF were revised based on experiences gained from the tests. By demonstrating satisfactory performance of the UVF, KHNP acquired an operational license in October, 2008 as an amendment to the operational license of the Ulchin NPPs. We are planning to conduct a simulated

  8. High-Level Waste Systems Plan. Revision 7 (U)

    International Nuclear Information System (INIS)

    Brooke, J.N.; Gregory, M.V.; Paul, P.; Taylor, G.; Wise, F.E.; Davis, N.R.; Wells, M.N.

    1996-10-01

    This revision of the High-Level Waste (HLW) System Plan aligns SRS HLW program planning with the DOE Savannah River (DOE-SR) Ten Year Plan (QC-96-0005, Draft 8/6), which was issued in July 1996. The objective of the Ten Year Plan is to complete cleanup at most nuclear sites within the next ten years. The two key principles of the Ten Year Plan are to accelerate the reduction of the most urgent risks to human health and the environment and to reduce mortgage costs. Accordingly, this System Plan describes the HLW program that will remove HLW from all 24 old-style tanks, and close 20 of those tanks, by 2006 with vitrification of all HLW by 2018. To achieve these goals, the DWPF canister production rate is projected to climb to 300 canisters per year starting in FY06, and remain at that rate through the end of the program in FY18, (Compare that to past System Plans, in which DWPF production peaked at 200 canisters per year, and the program did not complete until 2026.) An additional $247M (FY98 dollars) must be made available as requested over the ten year planning period, including a one-time $10M to enhance Late Wash attainment. If appropriate resources are made available, facility attainment issues are resolved and regulatory support is sufficient, then completion of the HLW program in 2018 would achieve a $3.3 billion cost savings to DOE, versus the cost of completing the program in 2026. Facility status information is current as of October 31, 1996

  9. Development of a combined soil-wash/in-furnace vitrification system for soil remediation at DOE sites

    International Nuclear Information System (INIS)

    Pegg, I.L.; Guo, Y.; Lahoda, E.J.; Lai, Shan-Tao; Muller, I.S.; Ruller, J.; Grant, D.C.

    1993-01-01

    This report addresses research and development of technologies for treatment of radioactive and hazardous waste streams at DOE sites. Weldon Spring raffinate sludges were used in a direct vitrification study to investigate their use as fluxing agents in glass formulations when blended with site soil. Storm sewer sediments from the Oak Ridge, TN, Y-12 facility were used for soil washing followed by vitrification of the concentrates. Both waste streams were extensively characterized. Testing showed that both mercury and uranium could be removed from the Y-12 soil by chemical extraction resulting in an 80% volume reduction. Thermal desorption was used on the contaminant-enriched minority fraction to separate the mercury from the uranium. Vitrification tests demonstrated that high waste loading glasses could be produced from the radioactive stream and from the Weldon Spring wastes which showed very good leach resistance, and viscosities and electrical conductivities in the range suitable for joule-heated ceramic melter (JHCM) processing. The conceptual process described combines soil washing, thermal desorption, and vitrification to produce clean soil (about 90% of the input waste stream), non-radioactive mercury, and a glass wasteform; the estimated processing costs for that system are about $260--$400/yd 3 . Results from continuous melter tests performed using Duratek's advanced JHCM (Duramelter) system are also presented. Since life cycle cost estimates are driven largely by volume reduction considerations, the large volume reductions possible with these multi-technology, blended waste stream approaches can produce a more leach resistant wasteform at a lower overall cost than alternative technologies such as cementation

  10. INNOVATIVE FOSSIL FUEL FIRED VITRIFICATION TECHNOLOGY FOR SOIL REMEDIATION. FINAL REPORT

    International Nuclear Information System (INIS)

    J. Hnat; L.M. Bartone; M. Pineda

    2001-01-01

    This Final Report summarizes the progress of Phases 3,3A and 4 of a waste technology Demonstration Project sponsored under a DOE Environmental Management Research and Development Program and administered by the U.S. Department of Energy National Energy Technology Laboratory-Morgantown (DOE-NETL) for an ''Innovative Fossil Fuel Fired Vitrification Technology for Soil Remediation''. The Summary Reports for Phases 1 and 2 of the Program were previously submitted to DOE. The total scope of Phase 3 was to have included the design, construction and demonstration of Vortec's integrated waste pretreatment and vitrification process for the treatment of low level waste (LLW), TSCA/LLW and mixed low-level waste (MLLW). Due to funding limitations and delays in the project resulting from a law suit filed by an environmental activist and the extended time for DOE to complete an Environmental Assessment for the project, the scope of the project was reduced to completing the design, construction and testing of the front end of the process which consists of the Material Handling and Waste Conditioning (MH/C) Subsystem of the vitrification plant. Activities completed under Phases 3A and 4 addressed completion of the engineering, design and documentation of the MH/C System such that final procurement of the remaining process assemblies can be completed and construction of a Limited Demonstration Project be initiated in the event DOE elects to proceed with the construction and demonstration testing of the MH/C Subsystem. Because of USEPA policies and regulations that do not require treatment of low level or low-level/PCB contaminated wastes, DOE terminated the project because there is no purported need for this technology

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

    International Nuclear Information System (INIS)

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

    2013-01-01

    The purpose of this Process Development Plan is to summarize the objectives and plans for the technology development activities for an alternative path for disposition of the recycle stream that will be generated in the Hanford Waste Treatment Plant Low Activity Waste (LAW) vitrification facility (LAW Recycle). This plan covers the first phase of the development activities. The baseline plan for disposition of this stream is to recycle it to the WTP Pretreatment Facility, where it will be concentrated by evaporation and returned to the LAW vitrification facility. Because this stream contains components that are volatile at melter temperatures and are also problematic for the glass waste form, they accumulate in the Recycle stream, exacerbating their impact on the number of LAW glass containers. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and reducing the halides in the Recycle is a key component of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, this stream does not have a proven disposition path, and resolving this gap becomes vitally important. This task seeks to examine the impact of potential future disposition of this stream in the Hanford tank farms, and to develop a process that will remove radionuclides from this stream and allow its diversion to another disposition path, greatly decreasing the LAW vitrification mission duration and quantity of glass waste. The origin of this LAW Recycle stream will be from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover or precipitates of scrubbed components (e.g. carbonates). The soluble

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-29

    The purpose of this Process Development Plan is to summarize the objectives and plans for the technology development activities for an alternative path for disposition of the recycle stream that will be generated in the Hanford Waste Treatment Plant Low Activity Waste (LAW) vitrification facility (LAW Recycle). This plan covers the first phase of the development activities. The baseline plan for disposition of this stream is to recycle it to the WTP Pretreatment Facility, where it will be concentrated by evaporation and returned to the LAW vitrification facility. Because this stream contains components that are volatile at melter temperatures and are also problematic for the glass waste form, they accumulate in the Recycle stream, exacerbating their impact on the number of LAW glass containers. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and reducing the halides in the Recycle is a key component of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, this stream does not have a proven disposition path, and resolving this gap becomes vitally important. This task seeks to examine the impact of potential future disposition of this stream in the Hanford tank farms, and to develop a process that will remove radionuclides from this stream and allow its diversion to another disposition path, greatly decreasing the LAW vitrification mission duration and quantity of glass waste. The origin of this LAW Recycle stream will be from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover or precipitates of scrubbed components (e.g. carbonates). The soluble

  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. 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

  15. Vitrification of plutonium at Rocky Flats the argument for a pilot plant

    Energy Technology Data Exchange (ETDEWEB)

    Moore, L. [Rocky Mountain Peace Center, Boulder, CO (United States)

    1996-05-01

    Current plans for stabilizing and storing the plutonium at Rocky Flats Plant fail to put the material in a form suitable for disposition and resistant to proliferation. Vitrification should be considered as an alternate technology. The vitrification should begin with a small-scale pilot plant.

  16. 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

  17. Commercial Ion Exchange Resin Vitrification Studies

    International Nuclear Information System (INIS)

    Cicero-Herman, C.A

    2002-01-01

    In the nuclear industry, ion exchange resins are used for purification of aqueous streams. The major contaminants of the resins are usually the radioactive materials that are removed from the aqueous streams. The use of the ion exchange resins creates a waste stream that can be very high in both organic and radioactive constituents. Therefore, disposal of the spent resin often becomes an economic problem because of the large volumes of resin produced and the relatively few technologies that are capable of economically stabilizing this waste. Vitrification of this waste stream presents a reasonable disposal alternative because of its inherent destruction capabilities, the volume reductions obtainable, and the durable product that it produces

  18. Remotely-Controlled Shear for Dismantling Highly Radioactive Tools In Rokkasho Vitrification Facility - 12204

    Energy Technology Data Exchange (ETDEWEB)

    Mitsui, Takashi; Sawa, Shusuke; Sadaki, Akira; Awano, Toshihiko [IHI Corporation, 1 Shin-Nakahara-cho, Isogo-ku, Yokohama, Kanagawa (Japan); Cole, Matt [S.A. Technology Inc, 3985 S. Lincoln Ave., Ste. 100, Loveland CO 80537 (United States); Miura, Yasuhiko; Ino, Tooru [Japan Nuclear Fuel Limited, 4-108, Aza Okitsuke, Oaza Obuchi, Rokkasho-Mura, Kamikita-gun, Aomori (Japan)

    2012-07-01

    A high-level liquid waste vitrification facility in the Japanese Rokkasho Reprocessing Plant (RRP) is right in the middle of hot commissioning tests toward starting operation in fall of 2012. In these tests, various tools were applied to address issues occurring in the vitrification cell. Because of these tools' unplanned placement in the cell it has been necessary to dismantle and dispose of them promptly. One of the tools requiring removal is a rod used in the glass melter to improve glass pouring. It is composed of a long rod made of Inconel 601 or 625 and has been highly contaminated. In order to dismantle these tools and to remotely put them in a designated waste basket, a custom electric shear machine was developed. It was installed in a dismantling area of the vitrification cell by remote cranes and manipulators and has been successfully operated. It can be remotely dismantled and placed in a waste basket for interim storage. This is a very good example of a successful deployment of a specialty remote tool in a hot cell environment. This paper also highlights how commissioning and operations are done in the Japanese Rokkasho Reprocessing Plant. (authors)

  19. 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

  20. 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

  1. 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

  2. Long-lived legacy: Managing high-level and transuranic waste at the DOE Nuclear Weapons Complex. Background paper

    International Nuclear Information System (INIS)

    1991-05-01

    The document focuses on high-level and transuranic waste at the DOE nuclear weapons complex. Reviews some of the critical areas and aspects of the DOE waste problem in order to provide data and further analysis of important issues. Partial contents, High-Level Waste Management at the DOE Weapons Complex, are as follows: High-Level Waste Management: Present and Planned; Amount and Distribution; Current and Potential Problems; Vitrification; Calcination; Alternative Waste Forms for the Idaho National Engineering Laboratory; Technologies for Pretreatment of High-Level Waste; Waste Minimization; Regulatory Framework; Definition of High-Level Waste; Repository Delays and Contingency Planning; Urgency of High-Level Tank Waste Treatment; Technologies for High-Level Waste Treatment; Rethinking the Waste Form and Package; Waste Form for the Idaho National Engineering Laboratory; Releases to the Atmosphere; Future of the PUREX Plant at Hanford; Waste Minimization; Tritium Production; International Cooperation; Scenarios for Future HLW Production. Partial contents of Chapter 2, Managing Transuranic Waste at the DOE Nuclear Weapons Complex, are as follows: Transuranic Waste at Department of Energy Sites; Amount and Distribution; Waste Management: Present and Planned; Current and Potential Problems; Three Technologies for Treating Retrievably Stored Transuranic Waste; In Situ Vitrification; The Applied Research, Development, Demonstration, Testing, and Evaluation Plan (RDDT ampersand E); Actinide Conversion (Transmutation); Waste Minimization; The Regulatory Framework; Definition of, and Standards for, Disposal of Transuranic Waste; Repository Delays; Alternative Storage and Disposal Strategies; Remediation of Buried Waste; The Waste Isolation Pilot Plant; Waste Minimization; Scenarios for Future Transuranic Waste Production; Conditions of No-Migration Determination

  3. 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

  4. Arc plasma incineration of surrogate radioactive wastes

    International Nuclear Information System (INIS)

    Girold, C.; Cartier, R.; Taupiac, J.P.; Vandensteendam, C.; Baronnet, J.M.

    1995-01-01

    The aim of this presentation is to demonstrate the feasibility to substitute a single plasma reactor, where the arc is transferred on a melt glass bath, for several steps in an existing nuclear technological wastes incinerator. The incineration of wastes, the produced gas treatment and the vitrification of ashes issued from waste incineration are the three simultaneous functions of this new kind of reactor. The three steps of the work are described: first, post-combustion in an oxygen plasma of gases generated from the waste pyrolysis, then, vitrification of ashes from the calcination of wastes in the transferred plasma furnace and finally, incineration/vitrification of wastes in the same furnace

  5. Resolution of regulatory issues facing the DOE in situ vitrification program

    International Nuclear Information System (INIS)

    Corathers, L.A.

    1992-03-01

    In situ vitrification (ISV) is being developed by researchers at the Pacific Northwest Laboratory (PNL), Idaho National Engineering Laboratory (INEL), and Oak Ridge National Laboratory (ORNL) as a technology for remediating soils, underground storage tank residuals, and buried materials that have been contaminated with hazardous, radioactive, and mixed wastes (i.e., wastes containing both radioactive and hazardous wastes) at US Department of Energy (DOE) facilities. The goal of the DOE ISV technology development program (i.e., the ISV Integrated Program) is to ensure that ISV is a workable technology for environmental restoration applications for DOE and other agencies. A DOE complex-wide plan was prepared during Fiscal Year 1991 to coordinate all levels of activities associated with the deployment of ISV. As part of this plan, a programmatic regulatory strategy was developed which focused on the federal environmental, health, safety, and nuclear regulations, including the US Environmental Protection Agency (EPA) and DOE regulations, believed to have the most significant near-term impact on the use of ISV as a remediation technology. The portion of the programmatic regulatory strategy addressing compliance with the Comprehensive Environmental Response, Compensation and Liability Act, as amended, and the Resource Conservation and Recovery Act, as amended, is presented in this paper

  6. DEWATERING TREATMENT SCALE-UP TESTING RESULTS OF HANFORD TANK WASTES

    International Nuclear Information System (INIS)

    TEDESCHI AR

    2008-01-01

    This report documents CH2M HILL Hanford Group Inc. (CH2M HILL) 2007 dryer testing results in Richland, WA at the AMEC Nuclear Ltd., GeoMelt Division (AMEC) Horn Rapids Test Site. It provides a discussion of scope and results to qualify the dryer system as a viable unit-operation in the continuing evaluation of the bulk vitrification process. A 10,000 liter (L) dryer/mixer was tested for supplemental treatment of Hanford tank low-activity wastes, drying and mixing a simulated non-radioactive salt solution with glass forming minerals. Testing validated the full scale equipment for producing dried product similar to smaller scale tests, and qualified the dryer system for a subsequent integrated dryer/vitrification test using the same simulant and glass formers. The dryer system is planned for installation at the Hanford tank farms to dry/mix radioactive waste for final treatment evaluation of the supplemental bulk vitrification process

  7. 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

  8. 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

  9. International technology exchange in support of the Defense Waste Processing Facility wasteform production

    International Nuclear Information System (INIS)

    Kitchen, B.G.

    1989-01-01

    The nearly completed Defense Waste Processing Facility (DWPF) is a Department of Energy (DOE) facility at the Savannah River Site that is designed to immobilize defense high level radioactive waste (HLW) by vitrification in borosilicate glass and containment in stainless steel canisters suitable for storage in the future DOE HLW repository. The DWPF is expected to start cold operation later this year (1990), and will be the first full scale vitrification facility operating in the United States, and the largest in the world. The DOE has been coordinating technology transfer and exchange on issues relating to HLW treatment and disposal through bi-lateral agreements with several nations. For the nearly fifteen years of the vitrification program at Savannah River Laboratory, over two hundred exchanges have been conducted with a dozen international agencies involving about five-hundred foreign national specialists. These international exchanges have been beneficial to the DOE's waste management efforts through confirmation of the choice of the waste form, enhanced understanding of melter operating phenomena, support for paths forward in political/regulatory arenas, confirmation of costs for waste form compliance programs, and establishing the need for enhancements of melter facility designs. This paper will compare designs and schedules of the international vitrification programs, and will discuss technical areas where the exchanges have provided data that have confirmed and aided US research and development efforts, impacted the design of the DWPF and guided the planning for regulatory interaction and product acceptance

  10. 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

  11. Multiphase, multi-electrode Joule heat computations for glass melter and in situ vitrification simulations

    International Nuclear Information System (INIS)

    Lowery, P.S.; Lessor, D.L.

    1991-02-01

    Waste glass melter and in situ vitrification (ISV) processes represent the combination of electrical thermal, and fluid flow phenomena to produce a stable waste-from product. Computational modeling of the thermal and fluid flow aspects of these processes provides a useful tool for assessing the potential performance of proposed system designs. These computations can be performed at a fraction of the cost of experiment. Consequently, computational modeling of vitrification systems can also provide and economical means for assessing the suitability of a proposed process application. The computational model described in this paper employs finite difference representations of the basic continuum conservation laws governing the thermal, fluid flow, and electrical aspects of the vitrification process -- i.e., conservation of mass, momentum, energy, and electrical charge. The resulting code is a member of the TEMPEST family of codes developed at the Pacific Northwest Laboratory (operated by Battelle for the US Department of Energy). This paper provides an overview of the numerical approach employed in TEMPEST. In addition, results from several TEMPEST simulations of sample waste glass melter and ISV processes are provided to illustrate the insights to be gained from computational modeling of these processes. 3 refs., 13 figs

  12. Letter report: Minor component study for low-level radioactive waste glasses

    International Nuclear Information System (INIS)

    Li, H.

    1996-03-01

    During the waste vitrification process, troublesome minor components in low-level radioactive waste streams could adversely affect either waste vitrification rate or melter life-time. Knowing the solubility limits for these minor components is important to determine pretreatment options for waste streams and glass formulation to prevent or to minimize these problems during the waste vitrification. A joint study between Pacific Northwest Laboratory and Rensselaer Polytechnic Institute has been conducted to determine minor component impacts in low-level nuclear waste glass

  13. 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

  14. High level waste management in France

    International Nuclear Information System (INIS)

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

    1979-01-01

    This paper deals with solidification of radioactive wastes. The vitrification process has been selected, the Marcoule vitrification plant and the operation are described. Prospects for vitrification of fission product solutions from La Hague reprocessing plant are given

  15. 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

  16. Commercial Ion Exchange Resin Vitrification in Borosilicate Glass

    International Nuclear Information System (INIS)

    Cicero-Herman, C.A.; Workman, P.; Poole, K.; Erich, D.; Harden, J.

    1998-05-01

    Bench-scale studies were performed to determine the feasibility of vitrification treatment of six resins representative of those used in the commercial nuclear industry. Each resin was successfully immobilized using the same proprietary borosilicate glass formulation. Waste loadings varied from 38 to 70 g of resin/100 g of glass produced depending on the particular resin, with volume reductions of 28 percent to 68 percent. The bench-scale results were used to perform a melter demonstration with one of the resins at the Clemson Environmental Technologies Laboratory (CETL). The resin used was a weakly acidic meth acrylic cation exchange resin. The vitrification process utilized represented a approximately 64 percent volume reduction. Glass characterization, radionuclide retention, offgas analyses, and system compatibility results will be discussed in this paper

  17. Certification plan transuranic waste: Hazardous Waste Handling Facility

    International Nuclear Information System (INIS)

    1992-06-01

    The purpose of this plan is to describe the organization and methodology for the certification of transuranic (TRU) waste handled in the Hazardous Waste Handling Facility at Lawrence Berkeley Laboratory (LBL). The 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 Quality Assurance Implementing Management Plan (QAIMP) for the HWBF; and a list of the current and planned implementing procedures used in waste certification

  18. Conceptual design for vitrification of HLW at West Valley using a rotary calciner/metallic melter

    International Nuclear Information System (INIS)

    Giraud, J.P.; Conord, J.P.; Saverot, P.M.

    1984-01-01

    The CEA has had an extensive research program in the field of vitrification technology for over 24 years, and several testing facilities were used throughout all phases of development and engineering: The Vulcain facility comprises a vitrification hot cell and four auxiliary hot cells. Vulcain allows the production of 2-kg samples of active glass. The off-gas treatment system allows testing the DF of each equipment. The auxiliary cells are equipped with leach-rate tests, diffusion tests, and irradiation tests on the glass samples. The Atlas facility is a reproduction of AVM calcination and vitrification furnaces at 1/2 scale enclosed in a glove box. This facility is used for testing ruthenium volatility and containment in the vitrification process. The full-scale AVM inactive pilot facility is used for testing calcination and vitrification of new compositions of high-level waste and for developing new types of vitrification furnaces. The inactive test loop is for testing air cooling of glass containers. The full-scale AVH inactive pilot facility is used for testing AVH technology and has been in operation since late 1981

  19. Stabilization of vitrified wastes: Task 4. Topical report, October 1994--September 1995

    International Nuclear Information System (INIS)

    Nowok, J.W.; Pflughoeft-Hassett, D.F.; Hassett, D.J.; Hurley, J.P.

    1995-09-01

    The goal of this task was to work with private industry to refine existing vitrification processes to produce a more stable vitrified product. The initial objectives were to (1) demonstrate a waste vitrification procedure for enhanced stabilization of waste materials and (2) develop a testing protocol to understand the long-term leaching behavior of the stabilized waste form. The testing protocol was expected to be based on a leaching procedure called the synthetic groundwater leaching procedure (SGLP). This task will contribute to the US DOE's identified technical needs in waste characterization, low-level mixed-waste processing, disposition technology, and improved waste forms. The proposed work was to proceed over 4 years in the following steps: literature surveys to aid in the selection and characterization of test mixtures for vitrification, characterization of optimized vitrified test wastes using advanced leaching protocols, and refinement and demonstration of vitrification methods leading to commercialization. For this year, literature surveys were completed, and computer modeling was performed to determine the feasibility of removing heavy metals from a waste during vitrification, thereby reducing the hazardous nature of the vitrified material and possibly producing a commercial metal concentrate. This report describes the following four subtasks: survey of vitrification technologies; survey of cleanup sites; selection and characterization of test mixtures for vitrification and crystallization; and selection of crystallization methods based on thermochemistry modeling

  20. 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.'

  1. Tank waste remediation system optimized processing strategy with an altered treatment scheme

    International Nuclear Information System (INIS)

    Slaathaug, E.J.

    1996-03-01

    This report provides an alternative strategy evolved from the current Hanford Site Tank Waste Remediation System (TWRS) programmatic baseline for accomplishing the treatment and disposal of the Hanford Site tank wastes. This optimized processing strategy with an altered treatment scheme performs the major elements of the TWRS Program, but modifies the deployment of selected treatment technologies to reduce the program cost. The present program for development of waste retrieval, pretreatment, and vitrification technologies continues, but the optimized processing strategy reuses a single facility to accomplish the separations/low-activity waste (LAW) vitrification and the high-level waste (HLW) vitrification processes sequentially, thereby eliminating the need for a separate HLW vitrification facility

  2. 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

  3. Waste Management Facilities Cost Information Report

    Energy Technology Data Exchange (ETDEWEB)

    Feizollahi, F.; Shropshire, D.

    1992-10-01

    The Waste Management Facility Cost Information (WMFCI) Report, commissioned by the US Department of Energy (DOE), develops planning life-cycle cost (PLCC) estimates for treatment, storage, and disposal facilities. This report contains PLCC estimates versus capacity for 26 different facility cost modules. A procedure to guide DOE and its contractor personnel in the use of estimating data is also provided. Estimates in the report apply to five distinctive waste streams: low-level waste, low-level mixed waste, alpha contaminated low-level waste, alpha contaminated low-level mixed waste, and transuranic waste. The report addresses five different treatment types: incineration, metal/melting and recovery, shredder/compaction, solidification, and vitrification. Data in this report allows the user to develop PLCC estimates for various waste management options.

  4. Waste Management Facilities Cost Information Report

    International Nuclear Information System (INIS)

    Feizollahi, F.; Shropshire, D.

    1992-10-01

    The Waste Management Facility Cost Information (WMFCI) Report, commissioned by the US Department of Energy (DOE), develops planning life-cycle cost (PLCC) estimates for treatment, storage, and disposal facilities. This report contains PLCC estimates versus capacity for 26 different facility cost modules. A procedure to guide DOE and its contractor personnel in the use of estimating data is also provided. Estimates in the report apply to five distinctive waste streams: low-level waste, low-level mixed waste, alpha contaminated low-level waste, alpha contaminated low-level mixed waste, and transuranic waste. The report addresses five different treatment types: incineration, metal/melting and recovery, shredder/compaction, solidification, and vitrification. Data in this report allows the user to develop PLCC estimates for various waste management options

  5. 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)

  6. Hanford Waste Management Plan, 1987

    International Nuclear Information System (INIS)

    1987-01-01

    The purpose of the Hanford Waste Management Plan (HWMP) is to provide an integrated plan for the safe storage, interim management, and disposal of existing waste sites and current and future waste streams at the Hanford Site. The emphasis of this plan is, however, on the disposal of Hanford Site waste. The plans presented in the HWMP are consistent with the preferred alternative which is based on consideration of comments received from the public and agencies on the draft Hanford Defense Waste Environmental Impact Statement (HDW-EIS). Low-level waste was not included in the draft HDW-EIS whereas it is included in this plan. The preferred alternative includes disposal of double-shell tank waste, retrievably stored and newly generated TRU waste, one pre-1970 TRU solid waste site near the Columbia River and encapsulated cesium and strontium waste

  7. Large-scale commercial applications of the in situ vitrification remediation technology

    International Nuclear Information System (INIS)

    Campbell, B.E.; Hansen, J.E.; McElroy, J.L.; Thompson, L.E.; Timmerman, C.L.

    1994-01-01

    The first large-scale commercial application of the innovative In Situ Vitrification (ISV) remediation technology was completed at the Parsons Chemical/ETM Enterprises Superfund site in Michigan State midyear 1994. This project involved treating 4,800 tons of pesticide and mercury-contaminated soil. The project also involved performance of the USEPA SITE Program demonstration test for the ISV technology. The Parsons project involved consolidation and staging of contaminated soil from widespread locations on and nearby the site. This paper presents a brief description of the ISV technology along with case-study type information on these two sites and the performance of the ISV technology on them. The paper also reviews other remediation projects where ISV has been identified as the/a preferred remedy, and where ISV is currently planned for use. These sites include soils contaminated with pesticides, dioxin, PCP, paint wastes, and a variety of heavy metals. This review of additional sites also includes a description of a planned radioactive mixed waste remediation project in Australia that contains large amounts of plutonium, uranium, lead, beryllium, and metallic and other debris buried in limestone and dolomitic soil burial pits. Initial test work has been completed on this application, and preparations are now underway for pilot testing in Australia. This project will demonstrate the applicability of the ISV technology to the challenging application of buried mixed wastes

  8. Glass and nuclear wastes

    International Nuclear Information System (INIS)

    Sombret, C.

    1982-10-01

    Glass shows interesting technical and economical properties for long term storage of solidified radioactive wastes by vitrification or embedding. Glass composition, vitrification processes, stability under irradiation, thermal stability and aqueous corrosion are studied [fr

  9. 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

  10. PROGRESS & CHALLENGES IN CLEANUP OF HANFORDS TANK WASTES

    Energy Technology Data Exchange (ETDEWEB)

    HEWITT, W.M.; SCHEPENS, R.

    2006-01-23

    The River Protection Project (RPP), which is managed by the Department of Energy (DOE) Office of River Protection (ORP), is highly complex from technical, regulatory, legal, political, and logistical perspectives and is the largest ongoing environmental cleanup project in the world. Over the past three years, ORP has made significant advances in its planning and execution of the cleanup of the Hartford tank wastes. The 149 single-shell tanks (SSTs), 28 double-shell tanks (DSTs), and 60 miscellaneous underground storage tanks (MUSTs) at Hanford contain approximately 200,000 m{sup 3} (53 million gallons) of mixed radioactive wastes, some of which dates back to the first days of the Manhattan Project. The plan for treating and disposing of the waste stored in large underground tanks is to: (1) retrieve the waste, (2) treat the waste to separate it into high-level (sludge) and low-activity (supernatant) fractions, (3) remove key radionuclides (e.g., Cs-137, Sr-90, actinides) from the low-activity fraction to the maximum extent technically and economically practical, (4) immobilize both the high-level and low-activity waste fractions by vitrification, (5) interim store the high-level waste fraction for ultimate disposal off-site at the federal HLW repository, (6) dispose the low-activity fraction on-site in the Integrated Disposal Facility (IDF), and (7) close the waste management areas consisting of tanks, ancillary equipment, soils, and facilities. Design and construction of the Waste Treatment and Immobilization Plant (WTP), the cornerstone of the RPP, has progressed substantially despite challenges arising from new seismic information for the WTP site. We have looked closely at the waste and aligned our treatment and disposal approaches with the waste characteristics. For example, approximately 11,000 m{sup 3} (2-3 million gallons) of metal sludges in twenty tanks were not created during spent nuclear fuel reprocessing and have low fission product concentrations. We

  11. Generalized Test Plan for the Vitrification of Simulated High-Level -Waste Calcine in the Idaho National Laboratory's Bench -Scale Cold Crucible Induction Melter

    International Nuclear Information System (INIS)

    Maio, Vince

    2011-01-01

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

  12. Ash from a pulp mill boiler--characterisation and vitrification.

    Science.gov (United States)

    Ribeiro, Ana S M; Monteiro, Regina C C; Davim, Erika J R; Fernandes, M Helena V

    2010-07-15

    The physical, chemical and mineralogical characterisation of the ash resulting from a pulp mill boiler was performed in order to investigate the valorisation of this waste material through the production of added-value glassy materials. The ash had a particle size distribution in the range 0.06-53 microm, and a high amount of SiO(2) (approximately 82 wt%), which was present as quartz. To favour the vitrification of the ash and to obtain a melt with an adequate viscosity to cast into a mould, different amounts of Na(2)O were added to act as fluxing agent. A batch with 80 wt% waste load melted at 1350 degrees C resulting in a homogeneous transparent green-coloured glass with good workability. The characterisation of the produced glass by differential thermal analysis and dilatometry showed that this glass presents a stable thermal behaviour. Standard leaching tests revealed that the concentration of heavy metals in the leaching solution was lower than those allowed by the Normative. As a conclusion, by vitrification of batch compositions with adequate waste load and additive content it is possible to produce an ash-based glass that may be used in similar applications as a conventional silicate glass inclusively as a building ecomaterial. 2010 Elsevier B.V. All rights reserved.

  13. 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

  14. Treatability study work plan for in situ vitrification of seepage pit 1 in Waste Area Grouping 7 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Spalding, B.P.

    1994-07-01

    A treatability study is described that encompasses the application of in situ vitrification (ISV) to at least two segments of Oak Ridge National Laboratory (ORNL) seepage pit 1 by the end of fiscal year 1995. This treatability study will establish the field-scale technical performance of ISV for (1) attaining the required depth, nominally 15 ft, to incorporate source contamination within and beneath the pits; (2) demonstrating field capability for the overlapping melt settings that are necessary to achieve fused melt segments; (3) demonstrating off-gas handling technology for accommodating and minimizing the volatilization of 137 Cs; (4) demonstrating adequate site characterization techniques to predict ISV melting kinetics, processing temperatures, and product durability; and (5) promoting public acceptance of ISV technology by demonstrating its safety, implementability, site impacts, and air emissions and by coordinating the treatability study within the regulatory closure process. The initial step of this treatability study will be to gather the required site characterization data about pit 1 so that the in situ vitrification can be effectively and safely planned. The second phase will be the field ISV operations at pit 1 employing at least two settings to achieve overlapping and fused melts. Such field operations are likely to require 6 to 8 weeks. Following termination of ISV melting operations at pit 1 and demobilization of portable ISV equipment and the off-gas hood, posttest characterization activities will begin

  15. In situ vitrification pilot-scale radioactive test

    International Nuclear Information System (INIS)

    Timmerman, C.L.; Oma, K.H.

    1984-10-01

    Pacific Northwest Laboratory (PNL) is developing in situ vitrification (ISV) as an in-place stabilization technique for selected liquid radioactive waste disposal sites. The process melts the wastes and surrounding soil to produce a durable glass and crystalline waste form. These ISV process development testing and evaluation studies are being conducted for the US Department of Energy. The results of an ISV pilot-scale test conducted in June of 1983 in which soils contaminated with actual radioactive transuranic and mixed fission product elements were vitrified are discussed. The primary objectives of the radioactive test were to: demonstrate containment and confinement of the radioactive material; verify equipment performance of the power and off-gas systems; identify losses to the off-gas system; and characterize the behavior of the radioactive material in the vitrified soil. The test successfully demonstrated the processing containment of radionuclides both within the vitrified mass and in the off-gas system. No environmental release of radioactive material was measured during testing operations. The vitrified soil had a greater than 99% retention of all radionuclides. Losses to the off-gas system varied from less than or equal to 0.03% for particulate materials (plutonium and strontium) to 0.8% for cesium which is a more volatile element. The off-gas system effectively contained both volatile and entrained radioactive materials. Analysis of the vitrified soil revealed that all radionuclides were distributed throughout the vitrified zone, some more uniformly than others. No migration of radionuclides outside the vitrification zone occurred, as indicated by analysis of soil samples from around the block. Previous waste form leaching studies indicate an acceptable durability of the ISV product. 8 references, 34 figures, 8 tables

  16. In situ vitrification pilot-scale radioactive test

    Energy Technology Data Exchange (ETDEWEB)

    Timmerman, C.L.; Oma, K.H.

    1984-10-01

    Pacific Northwest Laboratory (PNL) is developing in situ vitrification (ISV) as an in-place stabilization technique for selected liquid radioactive waste disposal sites. The process melts the wastes and surrounding soil to produce a durable glass and crystalline waste form. These ISV process development testing and evaluation studies are being conducted for the US Department of Energy. The results of an ISV pilot-scale test conducted in June of 1983 in which soils contaminated with actual radioactive transuranic and mixed fission product elements were vitrified are discussed. The primary objectives of the radioactive test were to: demonstrate containment and confinement of the radioactive material; verify equipment performance of the power and off-gas systems; identify losses to the off-gas system; and characterize the behavior of the radioactive material in the vitrified soil. The test successfully demonstrated the processing containment of radionuclides both within the vitrified mass and in the off-gas system. No environmental release of radioactive material was measured during testing operations. The vitrified soil had a greater than 99% retention of all radionuclides. Losses to the off-gas system varied from less than or equal to 0.03% for particulate materials (plutonium and strontium) to 0.8% for cesium which is a more volatile element. The off-gas system effectively contained both volatile and entrained radioactive materials. Analysis of the vitrified soil revealed that all radionuclides were distributed throughout the vitrified zone, some more uniformly than others. No migration of radionuclides outside the vitrification zone occurred, as indicated by analysis of soil samples from around the block. Previous waste form leaching studies indicate an acceptable durability of the ISV product. 8 references, 34 figures, 8 tables.

  17. 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

  18. 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.

  19. Certification Plan, low-level 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 low-level radioactive waste (LLW) handled in the Hazardous Waste Handling Facility (HWHF) at Lawrence Berkeley Laboratory (LBL). This plan also 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 and a list of the current and planned implementing procedures used in waste certification. This plan provides guidance from the HWHF to waste generators, waste handlers, and the Waste Certification Specialist to enable them to conduct their activities and carry out their responsibilities in a manner that complies with the requirements of WHC-WAC. Waste generators have the primary responsibility for the proper characterization of LLW. The Waste Certification Specialist verifies and certifies that LBL LLW is characterized, handled, and shipped in accordance with the requirements of WHC-WAC. Certification is the governing process in which LBL personnel conduct their waste generating and waste handling activities in such a manner that the Waste Certification Specialist can verify that the requirements of WHC-WAC are met

  20. In situ vitrification: Preliminary results from the first large-scale radioactive test

    International Nuclear Information System (INIS)

    Buelt, J.L.; Westsik, J.H.

    1988-02-01

    The first large-scale radioactive test (LSRT) of In Situ Vitrification (ISV) has been completed. In Situ Vitrification is a process whereby joule heating immobilizes contaminated soil in place by converting it to a durable glass and crystalline waste form. The LSRT was conducted at an actual transuranic contaminated soil site on the Department of Energy's Hanford Site. The test had two objectives: (1) determine large-scale processing performance and (2) produce a waste form that can be fully evaluated as a potential technique for the final disposal of transuranic-contaminated soil sites at Hanford. This accomplishment has provided technical data to evaluate the ISV process for its potential in the final disposition of transuranic-contaminated soil sites at Hanford. Because of the test's successful completion, within a year technical data on the vitrified soil will be available to determine how well the process incorporates transuranics into the waste form and how well the form resists leaching of transuranics. Preliminary results available include retention of transuranics and other elements within the waste form during processing and the efficiency of the off-gas treatment system in removing contaminants from the gaseous effluents. 13 refs., 10 figs., 5 tabs

  1. Vitrification of underground storage tanks: Technology development, regulatory issues, and cost analysis

    International Nuclear Information System (INIS)

    Tixier, J.S.; Corathers, L.A.; Anderson, L.D.

    1992-03-01

    In situ vitrification (ISV), developed by the Pacific Northwest Laboratory (PNL) for the US Department of Energy (DOE), is a thermal treatment process for the remediation of hazardous, radioactive, or mixed waste sites. The process has been broadly patented both domestically and abroad. Since the inception of ISV in 1980, developmental activities have been focused on applications to contaminated soils, and more recently the potential for application to buried wastes and underground structures (tanks). Research performed to date on the more advanced ISV applications (i.e., application to buried wastes and underground tanks) shows that significant technical and economic potential exists for using ISV to treat buried wastes and underground structures containing radionuclides and/or hazardous constituents. Present ISV applications are directed to the treatment of contaminated soils; the likelihood of using ISV to treat underground tanks depends on the resolution of significant technical and institutional issues related to this advanced application. This paper describes the ISV process and summarizes the technical progress of underground tank vitrification (UTV), discusses pertinent regulatory issues facing the use of UTV, and presents the potential cost of UTV relative to other remedial action alternatives

  2. 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.

  3. Characterization of vitrified soil produced by in situ vitrification

    International Nuclear Information System (INIS)

    Timmerman, C.L.; Lokken, R.O.

    1984-01-01

    Radioactive or other hazardous wastes buried at waste disposal sites may require further stabilization to secure the isolation of these wastes from the environment. One method of waste stabilization being developed is in situ vitrification. This process involves the in-place melting of buried wastes and the surrounding soil to produce a glass and crystalline waste form. Engineering-scale and pilot-scale demonstrations of this concept with soil contaminated with nonradioactive, hazardous species (Cs, Sr, Ru, Pb, Cd, etc.) were performed. These demonstrations provided information on species migration, crystalline-phase formation, and waste form durability. In addition to the nonradioactive tests, a crucible-scale melt of soil spiked with radioactive uranium, plutonium, and cesium was leach tested. The results show that hazardous waste components are retained in the product. The durability of the waste form in both the vitreous and the crystalline phases is similar to that of Pyrex glass

  4. Characterization of vitrified soil produced by in-situ vitrification

    International Nuclear Information System (INIS)

    Timmerman, C.L.; Lokken, R.O.

    1983-01-01

    Radioactive or other hazardous wastes buried at waste-disposal sites may require further stabilization to secure the isolation of these wastes from the environment. One method of waste stabilization being developed is in-situ vitrification. This process involves the in-place melting of buried wastes and the surrounding soil to produce a glass and crystalline waste form. Engineering-scale and pilot-scale demonstrations of this concept with soil contaminated with nonradioactive, hazardous species (Cs, Sr, Ru, Pb, Cd, etc.) were performed. These demonstrations provided information on species migration, crystalline phase formation, and waste form durability. In addition to the nonradioactive tests, a crucible-scale melt of soil spiked with radioactive uranium, plutonium, and cesium was leach tested. The results show that hazardous waste components are retained in the product. The durability of the waste form in both the vitreous and crystalline phases is similar to that of pyrex glass

  5. Facility effluent monitoring plan for the 324 Facility

    International Nuclear Information System (INIS)

    1994-11-01

    The 324 Facility [Waste Technology Engineering Laboratory] in the 300 Area primarily supports the research and development of radioactive and nonradioactive waste vitrification technologies, biological waste remediation technologies, spent nuclear fuel studies, waste mixing and transport studies, and tritium development programs. All of the above-mentioned programs deal with, and have the potential to, release hazardous and/or radioactive material. The potential for discharge would primarily result from (1) conducting research activities using the hazardous materials, (2) storing radionuclides and hazardous chemicals, and (3) waste accumulation and storage. This report summarizes the airborne and liquid effluents, and the results of the Facility Effluent Monitoring Plan (FEMP) determination for the facility. The complete monitoring plan includes characterizing effluent streams, monitoring/sampling design criteria, a description of the monitoring systems and sample analysis, and quality assurance requirements

  6. Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    SK Sundaram; ML Elliott; D Bickford

    1999-11-19

    SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described.

  7. Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

    International Nuclear Information System (INIS)

    Sundaram, S.K.; Elliott, M.L.; Bickford, D.

    1999-01-01

    SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described

  8. 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

  9. Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

  10. Comparison of simulants to actual neutralized current acid waste: Process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    Energy Technology Data Exchange (ETDEWEB)

    Morrey, E.V.; Tingey, J.M.

    1996-04-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs have been established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste is being performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

  11. Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    International Nuclear Information System (INIS)

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

    1996-10-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions

  12. Comparison of simulants to actual neutralized current acid waste: Process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    International Nuclear Information System (INIS)

    Morrey, E.V.; Tingey, J.M.

    1996-04-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs have been established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste is being performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions

  13. PROGRESS and CHALLENGES IN CLEANUP OF HANFORDS TANK WASTES

    International Nuclear Information System (INIS)

    HEWITT, W.M.; SCHEPENS, R.

    2006-01-01

    The River Protection Project (RPP), which is managed by the Department of Energy (DOE) Office of River Protection (ORP), is highly complex from technical, regulatory, legal, political, and logistical perspectives and is the largest ongoing environmental cleanup project in the world. Over the past three years, ORP has made significant advances in its planning and execution of the cleanup of the Hartford tank wastes. The 149 single-shell tanks (SSTs), 28 double-shell tanks (DSTs), and 60 miscellaneous underground storage tanks (MUSTs) at Hanford contain approximately 200,000 m 3 (53 million gallons) of mixed radioactive wastes, some of which dates back to the first days of the Manhattan Project. The plan for treating and disposing of the waste stored in large underground tanks is to: (1) retrieve the waste, (2) treat the waste to separate it into high-level (sludge) and low-activity (supernatant) fractions, (3) remove key radionuclides (e.g., Cs-137, Sr-90, actinides) from the low-activity fraction to the maximum extent technically and economically practical, (4) immobilize both the high-level and low-activity waste fractions by vitrification, (5) interim store the high-level waste fraction for ultimate disposal off-site at the federal HLW repository, (6) dispose the low-activity fraction on-site in the Integrated Disposal Facility (IDF), and (7) close the waste management areas consisting of tanks, ancillary equipment, soils, and facilities. Design and construction of the Waste Treatment and Immobilization Plant (WTP), the cornerstone of the RPP, has progressed substantially despite challenges arising from new seismic information for the WTP site. We have looked closely at the waste and aligned our treatment and disposal approaches with the waste characteristics. For example, approximately 11,000 m 3 (2-3 million gallons) of metal sludges in twenty tanks were not created during spent nuclear fuel reprocessing and have low fission product concentrations. We plan to

  14. The R and D and commercial experience on KHNP's vitrification technology

    International Nuclear Information System (INIS)

    Jo, Hyun-Jun; Kim, Cheon-Woo

    2015-01-01

    The Korea Hydro and Nuclear Power Co., Ltd., (KHNP) has investigated and evaluated various efficient thermal treatment technologies for the LILW. In 1994 and 1995, the feasibility of several melter technologies was assessed from technical and economic perspectives. Finally, the R and D project to develop the vitrification technology using CCIM (Cold Crucible Induction Melter) and PTM (Plasma Torch Melter) was launched in 1997. This R and D project had been completed from 1997 to 2002. KHNP started the project to construct the commercial facility using the results of the R and D project in 2002. The HanUl 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 CCIM technology. The design of UVF had been conducted from 2002 to 2005. The construction was begun in 2005 and was completed in 2007. From 2007 to 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 the commercial operation in October 2009. Based on the successful construction and operation of UVF, the advanced R and D project has been started to develop the large-scale vitrification facility. (author)

  15. Waste removal sequencing using ProdMod

    International Nuclear Information System (INIS)

    Paul, P.K.; Gregory, M.V.; Davis, N.R.; Brooke, J.N.

    1996-01-01

    The Savannah River Site (SRS) is starting to solidify its accumulated high-level radioactive waste into borosilicate glass in stainless steel canisters for eventual permanent storage. The in-tank precipitation process (ITP) and extended sludge processing (ESP) are two key operations in the waste processing complex. The supernate and dissolved salt from the waste storage tanks are transferred to the ITP process tank where the solution is decontaminated in batch processes. Soluble radioactive cesium is precipitated with sodium tetraphenylborate and strontium, uranium, and plutonium are adsorbed on monosodium titanate. The precipitate and adsorbent solids, which now contain the radionuclides, are concentrated using crossflow filters. The concentrated solids are sent to the high-level waste vitrification process. The decontaminated salt solution is sent to the low-level waste solidification process to form cement grout. In parallel with the precipitate operations, insoluble sludges that settled originally to the bottom of the waste tanks are reslurried and sent to ESP to undergo washing to reduce soluble salt content and aluminum dissolution, if required. In the vitrification process in the Defense Waste Processing Facility (DWPF), the concentrated precipitate from the ITP is mixed with the washed sludge from ESP and glass frit in proportion to form a stable borosilicate glass. A novel and fast-running Production Planning Model (ProdMod) has been developed to simulate the waste processing operation. This paper describes the application of ProdMod in sequencing the ITP batches and scheduling the ESP batches

  16. Statistical process control: An approach to quality assurance in the production of vitrified nuclear waste

    International Nuclear Information System (INIS)

    Pulsipher, B.A.; Kuhn, W.L.

    1987-01-01

    Current planning for liquid high-level nuclear wastes existing in the United States includes processing in a liquid-fed ceramic melter to incorporate it into a high-quality glass, and placement in a deep geologic repository. The nuclear waste vitrification process requires assurance of a quality product with little or no final inspection. Statistical process control (SPC) is a quantitative approach to one quality assurance aspect of vitrified nuclear waste. This method for monitoring and controlling a process in the presence of uncertainties provides a statistical basis for decisions concerning product quality improvement. Statistical process control is shown to be a feasible and beneficial tool to help the waste glass producers demonstrate that the vitrification process can be controlled sufficiently to produce an acceptable product. This quantitative aspect of quality assurance could be an effective means of establishing confidence in the claims to a quality product

  17. Statistical process control: An approach to quality assurance in the production of vitrified nuclear waste

    International Nuclear Information System (INIS)

    Pulsipher, B.A.; Kuhn, W.L.

    1987-02-01

    Current planning for liquid high-level nuclear wastes existing in the US includes processing in a liquid-fed ceramic melter to incorporate it into a high-quality glass, and placement in a deep geologic repository. The nuclear waste vitrification process requires assurance of a quality product with little or no final inspection. Statistical process control (SPC) is a quantitative approach to one quality assurance aspect of vitrified nuclear waste. This method for monitoring and controlling a process in the presence of uncertainties provides a statistical basis for decisions concerning product quality improvement. Statistical process control is shown to be a feasible and beneficial tool to help the waste glass producers demonstrate that the vitrification process can be controlled sufficiently to produce an acceptable product. This quantitative aspect of quality assurance could be an effective means of establishing confidence in the claims to a quality product. 2 refs., 4 figs

  18. 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.

  19. Product evaluation of in situ vitrification engineering, Test 4

    International Nuclear Information System (INIS)

    Loehr, C.A.; Weidner, J.R.; Bates, S.O.

    1991-09-01

    This report is one of several that evaluates the In Situ Vitrification (ISV) Engineering-Scale Test 4 (ES-4). This document describes the chemical and physical composition, microstructure, and leaching characteristics of ES-4 product samples; these data provide insight into the expected performance of a vitrified product in an ISV buried waste application similar to that studied in ES-4

  20. Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    1978-01-01

    A summary of the planning of transportation and plant design in the Swedish KBS project on management and disposal reprocessed radioactive waste. It describes a transportation system, a central storage facility for used fuel elements, a plant for intermediate storage and encapsulation and a final repository for the vitrified waste. Accounts are given for the reprocessing and vitrification. The safety of the entire system is discussed

  1. Towards optimization of nuclear waste glass: Constraints, property models, and waste loading

    International Nuclear Information System (INIS)

    Hrma, P.

    1994-04-01

    Vitrification of both low- and high-level wastes from 177 tanks at Hanford poses a great challenge to glass makers, whose task is to formulate a system of glasses that are acceptable to the federal repository for disposal. The enormous quantity of the waste requires a glass product of the lowest possible volume. The incomplete knowledge of waste composition, its variability, and lack of an appropriate vitrification technology further complicates this difficult task. A simple relationship between the waste loading and the waste glass volume is presented and applied to the predominantly refractory (usually high-activity) and predominantly alkaline (usually low-activity) waste types. Three factors that limit waste loading are discussed, namely product acceptability, melter processing, and model validity. Glass formulation and optimization problems are identified and a broader approach to uncertainties is suggested

  2. 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

  3. Second ILAW Site Borehole Characterization Plan

    International Nuclear Information System (INIS)

    Reidel, S.P.

    2000-01-01

    The US Department of Energy's Hanford Site has the most diverse and largest amounts of radioactive tank waste in the US. High-level radioactive waste has been stored at Hanford since 1944. Approximately 209,000 m 3 (54 Mgal) of waste are currently stored in 177 tanks. Vitrification and onsite disposal of low-activity tank waste (LAW) are embodied in the strategy described in the Tri-Party Agreement. The tank waste is to be retrieved, separated into low- and high-level fractions, and then immobilized. The low-activity vitrified waste will be disposed of in the 200 East Area of the Hanford Site. This report is a plan to drill and characterize the second borehole for the Performance Assessment. The first characterization borehole was drilled in 1998. The plan describes data collection activities for determining physical and chemical properties of the vadose zone and saturated zone on the northeast side of the proposed disposal site. These data will then be used in the 2005 Performance Assessment

  4. Treatability studies of actual listed waste sludges from the Oak Ridge Reservation (ORR)

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Peeler, D.K.; Gilliam, T.M.; Bleier, A.; Spence, R.D.

    1996-01-01

    Oak Ridge National Laboratory (ORNL) and Savannah River Technology Center (SRTC) are investigating vitrification for various low-level and mixed wastes on the Oak Ridge Reservation (ORR). Treatability studies have included surrogate waste formulations at the laboratory-, pilot-, and field-scales and actual waste testing at the laboratory- and pilot-scales. The initial waste to be processing through SRTC's Transportable Vitrification System (TVS) is the K-1407-B and K-1407-C (B/C) Pond sludge waste which is a RCRA F-listed waste. The B/C ponds at the ORR K-25 site were used as holding and settling ponds for various waste water treatment streams. Laboratory-, pilot-, and field- scale ''proof-of-principle'' demonstrations are providing needed operating parameters for the planned field-scale demonstration with actual B/C Pond sludge waste at ORR. This report discusses the applied systems approach to optimize glass compositions for this particular waste stream through laboratory-, pilot-, and field-scale studies with surrogate and actual B/C waste. These glass compositions will maximize glass durability and waste loading while optimizing melt properties which affect melter operation, such as melt viscosity and melter refractory corrosion. Maximum waste loadings minimize storage volume of the final waste form translating into considerable cost savings

  5. Iron Phosphate Glasses: An Alternative for Vitrifying Certain Nuclear Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Delbert E. Day; Chandra S. Ray; Cheol-Woon Kim

    2004-12-28

    Vitrification of nuclear waste in a glass is currently the preferred process for waste disposal. DOE currently approves only borosilicate (BS) type glasses for such purposes. However, many nuclear wastes, presently awaiting disposal, have complex and diverse chemical compositions, and often contain components that are poorly soluble or chemically incompatible in BS glasses. Such problematic wastes can be pre-processed and/or diluted to compensate for their incompatibility with a BS glass matrix, but both of these solutions increases the wasteform volume and the overall cost for vitrification. Direct vitrification using alternative glasses that utilize the major components already present in the waste is preferable, since it avoids pre-treating or diluting the waste, and, thus, minimizes the wasteform volume and overall cost.

  6. Iron Phosphate Glasses: An Alternative for Vitrifying Certain Nuclear Wastes

    International Nuclear Information System (INIS)

    Day, Delbert E.; Ray, Chandra S.; Cheol-Woon Kim

    2004-01-01

    Vitrification of nuclear waste in a glass is currently the preferred process for waste disposal. DOE currently approves only borosilicate (BS) type glasses for such purposes. However, many nuclear wastes, presently awaiting disposal, have complex and diverse chemical compositions, and often contain components that are poorly soluble or chemically incompatible in BS glasses. Such problematic wastes can be pre-processed and/or diluted to compensate for their incompatibility with a BS glass matrix, but both of these solutions increases the wasteform volume and the overall cost for vitrification. Direct vitrification using alternative glasses that utilize the major components already present in the waste is preferable, since it avoids pre-treating or diluting the waste, and, thus, minimizes the wasteform volume and overall cost

  7. 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

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

    International Nuclear Information System (INIS)

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

    1989-02-01

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

  9. Formulation and preparation of Hanford Waste Treatment Plant direct feed low activity waste Effluent Management Facility core simulant

    Energy Technology Data Exchange (ETDEWEB)

    McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, Charles A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL; Adamson, Duane J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL

    2016-05-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator in the Effluent Management Facility (EMF) and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Alternate disposition would also eliminate this stream from recycling within WTP when it begins operations and would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other problems such a recycle stream present. This LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures and are problematic for the glass waste form, such as halides and sulfate. Because this stream will recycle within WTP, these components accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfate in the recycled Condensate and is a key outcome of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of this task was to formulate and prepare a simulant of the LAW Melter

  10. Central Waste Complex (CWC) Waste Analysis Plan

    International Nuclear Information System (INIS)

    ELLEFSON, M.D.

    2000-01-01

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

  11. Minimum Additive Waste Stabilization (MAWS)

    International Nuclear Information System (INIS)

    1994-02-01

    In the Minimum Additive Waste Stabilization(MAWS) concept, actual waste streams are utilized as additive resources for vitrification, which may contain the basic components (glass formers and fluxes) for making a suitable glass or glassy slag. If too much glass former is present, then the melt viscosity or temperature will be too high for processing; while if there is too much flux, then the durability may suffer. Therefore, there are optimum combinations of these two important classes of constituents depending on the criteria required. The challenge is to combine these resources in such a way that minimizes the use of non-waste additives yet yields a processable and durable final waste form for disposal. The benefit to this approach is that the volume of the final waste form is minimized (waste loading maximized) since little or no additives are used and vitrification itself results in volume reduction through evaporation of water, combustion of organics, and compaction of the solids into a non-porous glass. This implies a significant reduction in disposal costs due to volume reduction alone, and minimizes future risks/costs due to the long term durability and leach resistance of glass. This is accomplished by using integrated systems that are both cost-effective and produce an environmentally sound waste form for disposal. individual component technologies may include: vitrification; thermal destruction; soil washing; gas scrubbing/filtration; and, ion-exchange wastewater treatment. The particular combination of technologies will depend on the waste streams to be treated. At the heart of MAWS is vitrification technology, which incorporates all primary and secondary waste streams into a final, long-term, stabilized glass wasteform. The integrated technology approach, and view of waste streams as resources, is innovative yet practical to cost effectively treat a broad range of DOE mixed and low-level wastes

  12. 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

  13. Am/Cm Vitrification Process: Vitrification Material Balance Calculations

    International Nuclear Information System (INIS)

    Smith, F.G.

    2000-01-01

    This report documents material balance calculations for the Americium/Curium vitrification process and describes the basis used to make the calculations. The material balance calculations reported here start with the solution produced by the Am/Cm pretreatment process as described in ``Material Balance Calculations for Am/Cm Pretreatment Process (U)'', SRT-AMC-99-0178 [1]. Following pretreatment, small batches of the product will be further treated with an additional oxalic acid precipitation and washing. The precipitate from each batch will then be charged to the Am/Cm melter with glass cullet and vitrified to produce the final product. The material balance calculations in this report are designed to provide projected compositions of the melter glass and off-gas streams. Except for decanted supernate collected from precipitation and precipitate washing, the flowsheet neglects side streams such as acid washes of empty tanks that would go directly to waste. Complete listings of the results of the material balance calculations are provided in the Appendices to this report

  14. Defense waste management plan

    International Nuclear Information System (INIS)

    1983-06-01

    Defense high-level waste (HLW) and defense transuranic (TRU) waste are in interim storage at three sites, namely: at the Savannah River Plant, in South Carolina; at the Hanford Reservation, in Washington; and at the Idaho National Engineering Laboratory, in Idaho. Defense TRU waste is also in interim storage at the Oak Ridge National Laboratory, in Tennessee; at the Los Alamos National Laboratory, in New Mexico; and at the Nevada Test Site, in Nevada. (Figure E-2). This document describes a workable approach for the permanent disposal of high-level and transuranic waste from atomic energy defense activities. The plan does not address the disposal of suspect waste which has been conservatively considered to be high-level or transuranic waste but which can be shown to be low-level waste. This material will be processed and disposed of in accordance with low-level waste practices. The primary goal of this program is to utilize or dispose of high-level and transuranic waste routinely, safely, and effectively. This goal will include the disposal of the backlog of stored defense waste. A Reference Plan for each of the sites describes the sequence of steps leading to permanent disposal. No technological breakthroughs are required to implement the reference plan. Not all final decisions concerning the activities described in this document have been made. These decisions will depend on: completion of the National Environmental Policy Act process, authorization and appropriation of funds, agreements with states as appropriate, and in some cases, the results of pilot plant experiments and operational experience. The major elements of the reference plan for permanent disposal of defense high-level and transuranic waste are summarized

  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. Nuclear waste. DOE's program to prepare high-level radioactive waste for final disposal

    International Nuclear Information System (INIS)

    Bannerman, Carl J.; Owens, Ronald M.; Dowd, Leonard L.; Herndobler, Christopher S.; Purvine, Nancy R.; Stenersen, Stanley G.

    1989-11-01

    In summary, as of December 1988, the four sites collectively stored about 95 million gallons of high-level waste in underground tanks and bins. Approximately 57 million gallons are stored at Hanford, 34 million gallons at Savannah River, 3 million gallons at INEL, and 6 million gallons at West Valley. The waste is in several forms, including liquid, sludge, and dry granular materials, that make it unsuitable for permanent storage in its current state at these locations. Leaks from the tanks, designed for temporary storage, can pose an environmental hazard to surrounding land and water for thousands of years. DOE expects that when its waste processes at Savannah River, West Valley, and Hanford become operational, the high-level radioactive waste stored at these sites will be blended with other materials to immobilize it by forming a glass-like substance. The glass form will minimize the risk of environmental damage and make the waste more acceptable for permanent disposal in a geologic repository. At INEL, DOE is still considering various other immobilization and permanent disposal approaches. In July 1989, DOE estimated that it would cost about $13 billion (in fiscal year 1988 dollars) to retrieve, process, immobilize, and store the high-level waste until it can be moved to a permanent disposal site: about $5.3 billion is expected to be spent at Savannah River, $0.9 billion at West Valley, $2.8 billion at Hanford, and $4.0 billion at INEL. DOE has started construction at Savannah River and West Valley for facilities that will be used to transform the waste into glass (a process known as vitrification). These sites have each encountered schedule delays, and one has encountered a significant cost increase over earlier estimates. More specifically, the Savannah River facility is scheduled to begin high-level waste vitrification in 1992; the West Valley project, based on a January 1989 estimate, is scheduled to begin high-level waste vitrification in 1996, about 8

  17. Bulk Vitrification Performance Enhancement: Refractory Lining Protection Against Molten Salt Penetration

    Energy Technology Data Exchange (ETDEWEB)

    Hrma, Pavel R.; Bagaasen, Larry M.; Schweiger, Michael J.; Evans, Michael B.; Smith, Benjamin T.; Arrigoni, Benjamin M.; Kim, Dong-Sang; Rodriguez, Carmen P.; Yokuda, Satoru T.; Matyas, Josef; Buchmiller, William C.; Gallegos, Autumn B.; Fluegel, Alexander

    2007-08-06

    Bulk vitrification (BV) is a process that heats a feed material that consists of glass-forming solids and dried low-activity waste (LAW) in a disposable refractory-lined metal box using electrical power supplied through carbon electrodes. The feed is heated to the point that the LAW decomposes and combines with the solids to generate a vitreous waste form. This study supports the BV design and operations by exploring various methods aimed at reducing the quantities of soluble Tc in the castable refractory block portion of the refractory lining, which limits the effectiveness of the final waste form.

  18. RIVER PROTECTION PROJECT SYSTEM PLAN

    International Nuclear Information System (INIS)

    Certa, P.J.; Kirkbride, R.A.; Hohl, T.M.; Empey, P.A.; Wells, M.N.

    2009-01-01

    The U.S. Department of Energy (DOE), Office of River Protection (ORP) manages the River Protection Project (RPP). The RPP mission is to retrieve and treat Hanford's tank waste and close the tank farms to protect the Columbia River. As a result, ORP is responsible for the retrieval, treatment, and disposal of approximately 57 million gallons 1 of radioactive waste contained in the Hanford Site waste tanks and closure2 of all the tanks and associated facilities. The previous revision of the System Plan was issued in May 2008. ORP has made a number of changes to the tank waste treatment strategy and plans since the last revision of this document, and additional changes are under consideration. ORP has contracts in place to implement the strategy for completion of the mission and establish the capability to complete the overall mission. The current strategl involves a number of interrelated activities. ORP will reduce risk to the environment posed by tank wastes by the following: (1) Retrieving the waste from the single-shell tanks (SST) to double-shell tanks (DST) and delivering the waste to the Waste Treatment and Immobilization Plant (WTP). (2) Constructing and operating the WTP, which will safely treat all of the high-level waste (HLW) fraction contained in the tank farms. About one-third of the low-activity waste (LAW) fraction separated from the HLW fraction in the WTP will be immobilized in the WTP LAW Vitrification Facility. (3) Developing and deploying supplemental treatment capability assumed to be a second LAW vitrification facility that can safely treat about two-thirds of the LAW contained in the tank farms. (4) Developing and deploying supplemental pretreatment capability currently assumed to be an Aluminum Removal Facility (ARF) using a lithium hydrotalcite process to mitigate sodium management issues. (5) Developing and deploying treatment and packaging capability for contact-handled transuranic (CH-TRU) tank waste for possible shipment to and disposal

  19. RIVER PROTECTION PROJECT SYSTEM PLAN

    Energy Technology Data Exchange (ETDEWEB)

    CERTA PJ; KIRKBRIDE RA; HOHL TM; EMPEY PA; WELLS MN

    2009-09-15

    The U.S. Department of Energy (DOE), Office of River Protection (ORP) manages the River Protection Project (RPP). The RPP mission is to retrieve and treat Hanford's tank waste and close the tank farms to protect the Columbia River. As a result, ORP is responsible for the retrieval, treatment, and disposal of approximately 57 million gallons 1 of radioactive waste contained in the Hanford Site waste tanks and closure2 of all the tanks and associated facilities. The previous revision of the System Plan was issued in May 2008. ORP has made a number of changes to the tank waste treatment strategy and plans since the last revision of this document, and additional changes are under consideration. ORP has contracts in place to implement the strategy for completion of the mission and establish the capability to complete the overall mission. The current strategl involves a number of interrelated activities. ORP will reduce risk to the environment posed by tank wastes by the following: (1) Retrieving the waste from the single-shell tanks (SST) to double-shell tanks (DST) and delivering the waste to the Waste Treatment and Immobilization Plant (WTP). (2) Constructing and operating the WTP, which will safely treat all of the high-level waste (HLW) fraction contained in the tank farms. About one-third of the low-activity waste (LAW) fraction separated from the HLW fraction in the WTP will be immobilized in the WTP LAW Vitrification Facility. (3) Developing and deploying supplemental treatment capability assumed to be a second LAW vitrification facility that can safely treat about two-thirds of the LAW contained in the tank farms. (4) Developing and deploying supplemental pretreatment capability currently assumed to be an Aluminum Removal Facility (ARF) using a lithium hydrotalcite process to mitigate sodium management issues. (5) Developing and deploying treatment and packaging capability for contact-handled transuranic (CH-TRU) tank waste for possible shipment to and

  20. 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

  1. AVLIS production plant waste management plan

    International Nuclear Information System (INIS)

    1984-01-01

    Following the executive summary, this document contains the following: (1) waste management facilities design objectives; (2) AVLIS production plant wastes; (3) waste management design criteria; (4) waste management plan description; and (5) waste management plan implementation. 17 figures, 18 tables

  2. West Valley Demonstration Project vitrification process equipment Functional and Checkout Testing of Systems (FACTS)

    International Nuclear Information System (INIS)

    Carl, D.E.; Paul, J.; Foran, J.M.; Brooks, R.

    1990-01-01

    The Vitrification Facility (VF) at the West Valley Demonstration Project was designed to convert stored radioactive waste into a stable glass for disposal in a federal repository. The Functional and Checkout Testing of Systems (FACTS) program was conducted from 1984 to 1989. During this time new equipment and processes were developed, installed, and implemented. Thirty-seven FACTS tests were conducted, and approximately 150,000 kg of glass were made by using nonradioactive materials to simulate the radioactive waste. By contrast, the planned radioactive operation is expected to produce approximately 500,000 kg of glass. The FACTS program demonstrated the effectiveness of equipment and procedures in the vitrification system, and the ability of the VF to produce quality glass on schedule. FACTS testing also provided data to validate the WVNS waste glass qualification method and verify that the product glass would meet federal repository acceptance requirements. The system was built and performed to standards which would have enabled it to be used in radioactive service. As a result, much of the VF tested, such as the civil construction, feed mixing and holding vessels, and the off-gas scrubber, will be converted for radioactive operation. The melter was still in good condition after being at temperature for fifty-eight of the sixty months of FACTS. However, the melter exceeded its recommended design life and will be replaced with a similar melter. Components that were not designed for remote operation and maintenance will be replaced with remote-use items. The FACTS testing was accomplished with no significant worker injury or environmental releases. During the last FACTS run, the VF processes approximated the remote-handling system that will be used in radioactive operations. Following this run the VF was disassembled for conversion to a radioactive process. Functional and checkout testing of new components will be performed prior to radioactive operation

  3. Minor component study for simulated high-level nuclear waste glasses (Draft)

    International Nuclear Information System (INIS)

    Li, H.; Langowskim, M.H.; Hrma, P.R.; Schweiger, M.J.; Vienna, J.D.; Smith, D.E.

    1996-02-01

    Hanford Site single-shell tank (SSI) and double-shell tank (DSI) wastes are planned to be separated into low activity (or low-level waste, LLW) and high activity (or high-level waste, HLW) fractions, and to be vitrified for disposal. Formulation of HLW glass must comply with glass processibility and durability requirements, including constraints on melt viscosity, electrical conductivity, liquidus temperature, tendency for phase segregation on the molten glass surface, and chemical durability of the final waste form. A wide variety of HLW compositions are expected to be vitrified. In addition these wastes will likely vary in composition from current estimates. High concentrations of certain troublesome components, such as sulfate, phosphate, and chrome, raise concerns about their potential hinderance to the waste vitrification process. For example, phosphate segregation in the cold cap (the layer of feed on top of the glass melt) in a Joule-heated melter may inhibit the melting process (Bunnell, 1988). This has been reported during a pilot-scale ceramic melter run, PSCM-19, (Perez, 1985). Molten salt segregation of either sulfate or chromate is also hazardous to the waste vitrification process. Excessive (Cr, Fe, Mn, Ni) spinel crystal formation in molten glass can also be detrimental to melter operation

  4. Estimation of potential ecological hazard of solidificated waste disposal

    International Nuclear Information System (INIS)

    Krylova, N.V.

    1980-01-01

    The results of estimation of potential ecological hazard of vitrificated high-level radioactive wastes resulted from spent fuel reprocessing of LWR connected with a hypothetic storage damage being occurred in the 5O0-6000-year geologic period are presented. The total volume of the vitrificated wastes in the storage used for calculations is 12000 blocks. The data on vitrificated block radioactivity depending on the time after fuel regeneration, the density of the uniform distribution of vitrificated wastes over the earth surface, as well as the results of estimation of the man external and internal exposures due to radionuclide escape into the biosphere are given in tables. It is shown that the main hazard is caused by external irradiation. The inhalation dose may be significant for man, though the hazard due to radionuclide intake by ingestion is less

  5. The ANSTO waste management action plan

    International Nuclear Information System (INIS)

    Levins, D.

    1997-01-01

    ANSTO's Waste Management Action Plan is a five-year program which addresses legacy issues that have arisen from the accumulation of radioactive wastes at Lucas Heights over the last forty years. Following an extensive review of waste management practices, a detailed Action Plan was prepared involving seventeen projects in the areas of solid wastes, liquid wastes, control of effluents and emissions, spent reactor fuel and organisational issues. The first year of the Waste Management Action Plan has resulted in significant achievements, especially in the areas of improved storage of solid wastes, stabilisation of uranium scrap, commissioning and operation of a scanning system for low-level waste drums, treatment of intermediate-level liquid wastes and improvements in the methods for monitoring of spent fuel storage facilities. The main goal of the Waste Management Action Plan is to achieve consistency, by the year 2000, with best practice as identified in the Radioactive Waste Safety Standards and Guidelines currently under development by the IAEA

  6. PFP dangerous waste training plan

    International Nuclear Information System (INIS)

    Khojandi, J.

    1996-01-01

    This document establishes the minimum training requirements for the Plutonium Finishing Plant (PFP) personnel who are responsible for management of dangerous waste. The training plan outlines training requirements for handling of solid dangerous waste during generator accumulation and liquid dangerous waste during treatment and storage operations. The implementation of this training plan will ensure the PFP facility compliance with the training plan requirements of Dangerous Waste Regulation. Chapter 173-303-330. Washington Administrative Code (WAC). The requirements for such compliance is described in Section 11.0 of WHC-CM-7-5 Environmental Compliance Manual

  7. Scaling considerations for modeling the in situ vitrification process

    International Nuclear Information System (INIS)

    Langerman, M.A.; MacKinnon, R.J.

    1990-09-01

    Scaling relationships for modeling the in situ vitrification waste remediation process are documented based upon similarity considerations derived from fundamental principles. Requirements for maintaining temperature and electric potential field similarity between the model and the prototype are determined as well as requirements for maintaining similarity in off-gas generation rates. A scaling rationale for designing reduced-scale experiments is presented and the results are assessed numerically. 9 refs., 6 figs

  8. Tank waste remediation system retrieval and disposal mission readiness-to-proceed guidance and requirements to deliverables crosswalk

    International Nuclear Information System (INIS)

    Hall, C.E.

    1998-01-01

    Before RL can authorize proceeding with Phase 1B, the PHMC team must demonstrate its readiness to retrieve and deliver the waste to the private contractors and to receive and dispose of the products and byproducts returned from the treatment. The PHMC team has organized their plans for providing these vitrification-support services into the Retrieval and Disposal Mission within the Tank Waste Remediation System (TWRS) Program

  9. Off-gas considerations for a vitrification plant in the republic of Korea

    International Nuclear Information System (INIS)

    Chun, Ung Kyung; Park, Jong Kil; Yang, Kyung Hwa; Song, Myung Jae

    1997-01-01

    The Republic of Korea is in the process of preparing for its first ever vitrification plant to handle low and intermediate-level radioactive waste from her pressurized water reactors (PWRs). KEPRI, in coordination with her partners, will design, construct, and erect a pilot plant using data from the orientation tests. The pilot plant will be the basis for the development of the final objective, the establishment of an industrial scale vitrification installation in the Republic of Korea. Throughout these projects, the major goal is to minimize the harmful effects of the final waste form to the environment. The gaseous effluents emissions from the facility will need to be managed to meet the environmental regulations concerning gaseous releases into the environment of the Republic of Korea. The focus of this paper is on the considerations for the treatment of the off-gas for a low and intermediate-level radioactive waste treatment vitrification installation in the Republic of Korea. Off-gas considerations will span a wide-range of areas such as waste characteristics, thermal treatment systems, off-gas regulations, off-gas characteristics, assessment of air pollution control devices, systems assessments, numerical modelling, economics etc. Off-gas regulations in Korea are becoming tighter and will likely change from year to year. In terms of both off-gas treatment equipment performance and public protection, the amount and nature (e.g. chemical behavior and morphology) of the species are important. The emissions may be classified as toxic metals, radionuclides, hydrocarbons, particulate matter, and acid gases. Air pollution control technologies are generally classified as wet or dry technologies covering over 40 different air pollution control devices (APCDs) with varying removal efficiencies for the different types of off-gas. In general, the state of the art systems for vitrification technologies incorporate the basic functions such as further oxidation of products

  10. 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

  11. River Protection Project (RPP) Immobilized Low- Ativity Waste (ILAW) Disposal Plan

    International Nuclear Information System (INIS)

    BRIGGS, M.G.

    2000-01-01

    This document replaces HNF-1517, Rev 2 which is deleted. It incorporates updates to reflect changes in programmatic direction associated with the vitrification plant contract change and associated DOE/ORP guidance. In addition it incorporates the cancellation of Project W-465, Grout Facility, and the associated modifications to Project W-520, Immobilized High-Level Waste Disposal Facility. It also includes document format changes and section number modifications consistent with CH2M HILL Hanford Group, Inc. procedures

  12. Commercial waste treatment program annual progress report for FY 1983

    Energy Technology Data Exchange (ETDEWEB)

    McElroy, J.L.; Burkholder, H.C. (comps.)

    1984-02-01

    This annual report describes progress during FY 1983 relating to technologies under development by the Commercial Waste Treatment Program, including: development of glass waste form and vitrification equipment for high-level wastes (HLW); waste form development and process selection for transuranic (TRU) wastes; pilot-scale operation of a radioactive liquid-fed ceramic melter (LFCM) system for verifying the reliability of the reference HLW treatment proces technology; evaluation of treatment requirements for spent fuel as a waste form; second-generation waste form development for HLW; and vitrification process control and product quality assurance technologies.

  13. Cement encapsulation of low-level waste liquids. Final report

    International Nuclear Information System (INIS)

    Baker, M.N.; Houston, H.M.

    1999-01-01

    Pretreatment of liquid high-level radioactive waste at the West Valley Demonstration Project (WVDP) was essential to ensuring the success of high-level waste (HLW) vitrification. By chemically separating the HLW from liquid waste, it was possible to achieve a significant reduction in the volume of HLW to be vitrified. In addition, pretreatment made it possible to remove sulfates, which posed several processing problems, from the HLW before vitrification took place

  14. Assessment of LANL waste management site plan

    International Nuclear Information System (INIS)

    Black, R.L.; Davis, K.D.; Hoevemeyer, S.S.; Jennrich, E.A.; Lund, D.M.

    1991-04-01

    The objective of this report is to present findings from evaluating the Los Alamos National Laboratory (LANL) Waste Management Plan to determine if it meets applicable DOE requirements. DOE Order 5820.2A, Radioactive Waste Management, sets forth requirements and guidelines for the establishment of a Waste Management Plan. The primary purpose of a Waste Management Plan is to describe how waste operations are conducted, what facilities are being used to manage wastes, what forces are acting to change current waste management systems, and what plans are in store for the coming year

  15. In-situ vitrification of radioactively contaminated soils: summary paper

    International Nuclear Information System (INIS)

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

    1987-01-01

    The in-situ vitrification (ISV) process is a new technology that has been developed from its conceptual phase through selected field-scale application tests during the last six years. In situ vitrification converts contaminated soils and waste inclusions into a durable glass and crystalline waste form by in-place melting. Electrodes are inserted into the soil to be treated and an electrical current is passed through the soil to be treated and an electrical current is passed through the soil to melt it. After cooling, the process fixes (TRU) and fission product radionuclides making them relatively nonleachable, resistant to intrusion, and nondispersible when intentionally disturbed. Another application considered for isolation of radioactively contaminated soils, but not yet developed, is the generation of impermeable barrier walls to prevent ground water seepage into a site. The barrier technique could also be used over the surface of an existing disposal site to deter plant and animal intrusion. The development units have been extensively tested with many types of soils and waste inclusions such as concrete, buried metals, sealed containers, organic chemicals with high boiling points such as polychlorinated biphenyls, and inorganic chemicals, including toxic heavy metals, nitrates, and sulfates. Nitrates and organics are destroyed, while heavy metals and fluorides are retained to a high percentage within the molten soil during processing. At $200 to $300/m 3 for radioactive waste, the process is economically competitive with many alternative remediation processes. The ISV process has been developed to the point where it is ready for large-scale field testing at an actual TRU-contaminated soil site. 5 references, 2 figures, 2 tables

  16. Hazardous Waste Cerification Plan: Hazardous Waste Handling Facility, Lawrence Berkeley Laboratory

    International Nuclear Information System (INIS)

    1992-02-01

    The purpose of this plan is to describe the organization and methodology for the certification of hazardous waste (HW) handled in the Lawrence Berkeley Laboratory (LBL) Hazardous Waste Handling Facility (HWHF). The plan also 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; and executive summary of the Quality Assurance Program Plan (QAPP) for the HWHF and a list of the current and planned implementing procedures used in waste certification. The plan provides guidance from the HWHF to waste generators, waste handlers, and the Systems Group Manager to enable them to conduct their activities and carry out their responsibilities in a manner that complies with several requirements of the Federal Resource Conservation and Resource Recovery Act (RCRA), the Federal Department of Transportation (DOT), and the State of California, Code of Regulations (CCR), Title 22

  17. Tracer-level radioactive pilot-scale test of in situ vitrification technology for the stabilization of contaminated soil sites at ORNL

    International Nuclear Information System (INIS)

    Jacobs, G.K.; Spalding, B.P.

    1991-11-01

    This plan summarizes the activities to be performed during FY 1990 and FY 1991 for the tracer-level radioactive pilot-scale in situ vitrification (ISV) test. This test is the second step in evaluating ISV as a remedial action for the pits and trenches at Oak Ridge National Laboratory (ORNL). A previous test used nonradioactive tracers for cesium and strontium. This new test will again use a one-half-scale model of trench 7 and the pilot-scale ISV equipment of Pacific Northwest Laboratory (PNL). A small and precisely known amount of waste from a liquid waste disposal pit will be used for the test. An actually contaminated waste site cannot be used for this test because of the necessity to use an exactly known inventory of radionuclides so that a precise measurement of the volatilization of various constituents to the off-gas can be determined

  18. 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)

  19. Low-level waste certification plan

    International Nuclear Information System (INIS)

    Greenhalph, W.O.

    1995-01-01

    This plan describes the organization and methodology for the certification of solid low-level waste (LLW) and mixed-waste (MW) generated at any of the facilities or major work activities of the Engineered Process Application (EPA) organization. The primary LLW and MW waste generating facility operated by EPA is the 377 Building. This plan does not cover the handling of hazardous or non-regulated waste, though they are mentioned at times for completeness

  20. In-situ vitrification: a status of the technology

    International Nuclear Information System (INIS)

    FitzPatrick, V.F.

    1986-09-01

    The In Situ Vitrification (ISV) process is a new technology developed from its conceptual phase to selected field-scale applications in the last 5 years. The US Department of Energy (DOE) has sponsored the ISV program to develop alternative technology for potential application to contaminated soil sites. The ISV process converts contaminated soils and wastes into a durable glass and crystalline waste form in place by melting using joule heating. The ISV process has been developed through a series of 25 engineering-scale (laboratory) tests, 10 pilot-scale (small field) tests, and four large-scale (full-scale field) tests. Its major advantages for stabilizing radioactive and hazardous wastes are found to be: safety in terms of minimizing worker and public exposure; long-term durability of waste form (more than 1 million years); cost effectiveness ($150 to $300/m 3 ); applicability to a wide variety of soils and inclusions; and potential for eliminating exhumation, transport, and handling

  1. 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

  2. Solid Waste Management Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, D.R.

    1990-08-01

    The objective of the Solid Waste Management Program Plan (SWMPP) is to provide a summary level comprehensive approach for the storage, treatment, and disposal of current and future solid waste received at the Hanford Site (from onsite and offsite generators) in a manner compliant with current and evolving regulations and orders (federal, state, and Westinghouse Hanford Company (Westinghouse Hanford)). The Plan also presents activities required for disposal of selected wastes currently in retrievable storage. The SWMPP provides a central focus for the description and control of cost, scope, and schedule of Hanford Site solid waste activities, and provides a vehicle for ready communication of the scope of those activities to onsite and offsite organizations. This Plan represents the most complete description available of Hanford Site Solid Waste Management (SWM) activities and the interfaces between those activities. It will be updated annually to reflect changes in plans due to evolving regulatory requirements and/or the SWM mission. 8 refs., 9 figs., 4 tabs.

  3. Innovative vitrification for soil remediation

    Energy Technology Data Exchange (ETDEWEB)

    Jetta, N.W.; Patten, J.S.; Hnat, J.G. [Vortec Corp., Collegeville, PA (United States)

    1995-10-01

    The objective of this DOE demonstration program is to validate the performance and operation of the Vortec Cyclone Melting System (CMS{trademark}) for the processing of LLW contaminated soils found at DOE sites. This DOE vitrification demonstration project has successfully progressed through the first two phases. Phase I consisted of pilot scale testing with surrogate wastes and the conceptual design of a process plant operating at a generic DOE site. The objective of Phase 2, which is scheduled to be completed the end of FY 95, is to develop a definitive process plant design for the treatment of wastes at a specific DOE facility. During Phase 2, a site specific design was developed for the processing of LLW soils and muds containing TSCA organics and RCRA metal contaminants. Phase 3 will consist of a full scale demonstration at the DOE gaseous diffusion plant located in Paducah, KY. Several DOE sites were evaluated for potential application of the technology. Paducah was selected for the demonstration program because of their urgent waste remediation needs as well as their strong management and cost sharing financial support for the project.

  4. Vitrification melter study

    International Nuclear Information System (INIS)

    Jones, J.A.

    1995-04-01

    This report presents the results of a study performed to identify the most promising vitrification melter technologies that the Department of Energy (EM-50) might pursue with available funding. The primary focus was on plasma arc systems and graphite arc melters. The study was also intended to assist EM-50 in evaluating competing technologies, formulating effective technology strategy, developing focused technology development projects, and directing the work of contractors involved in vitrification melter development

  5. Nuclear waste management

    International Nuclear Information System (INIS)

    1982-12-01

    The subject is discussed, with special reference to the UK, under the headings: radiation; origins of the waste (mainly from nuclear power programme; gas, liquid, solid; various levels of activity); dealing with waste (methods of processing, storage, disposal); high-active waste (storage, vitrification, study of means of eventual disposal); waste management (UK organisation to manage low and intermediate level waste). (U.K.)

  6. Feed Materials Production Center Waste Management Plan

    International Nuclear Information System (INIS)

    Watts, R.E.; Allen, T.; Castle, S.A.; Hopper, J.P.; Oelrich, R.L.

    1986-01-01

    In the process of producing uranium metal products used in Department of Energy (DOE) defense programs at other DOE facilities, various types of wastes are generated at the Feed Materials Production Center (FMPC). Process wastes, both generated and stored, are discussed in the Waste Management Plan and include low-level radioactive waste (LLW), mixed hazardous/radioactive waste, and sanitary/industrial waste. Scrap metal waste and wastes requiring special remediation are also addressed in the Plan. The Waste Management Plan identifies the comprehensive programs developed to address safe storage and disposition of all wastes from past, present, and future operations at the FMPC. Waste streams discussed in this Plan are representative of the waste generated and waste types that concern worker and public health and safety. Budgets and schedules for implementation of waste disposition are also addressed. The waste streams receiving the largest amount of funding include LLW approved for shipment by DOE/ORO to the Nevada Test Site (NTS) (MgF 2 , slag leach filter cake, and neutralized raffinate); remedial action wastes (waste pits, K-65 silo waste); thorium; scrap metal (contaminated and noncontaminated ferrous and copper scrap); construction rubble and soil generated from decontamination and decommissioning of outdated facilities; and low-level wastes that will be handled through the Low-Level Waste Processing and Shipping System (LLWPSS). Waste Management milestones are also provided. The Waste Management Plan is divided into eight major sections: Introduction; Site Waste and Waste Generating Process; Strategy; Projects and Operations; Waste Stream Budgets; Milestones; Quality Assurance for Waste Management; and Environmental Monitoring Program

  7. Radiation exposure control by estimation of multiplication factors for online remote radiation monitoring systems at Vitrification Plant

    International Nuclear Information System (INIS)

    Deokar, Umesh V.; Kukarni, V.V.; Khot, A.R.; Mathew, P.; Kamlesh; Purohit, R.G.; Sarkar, P.K.

    2011-01-01

    Vitrification Plant is commissioned for vitrification of high-level liquid waste generated in Nuclear Fuel Cycle operations by using Joule Heated Ceramic Melter first time in India. Exposure control is a major concern in operating plant. Therefore, in addition to installed monitors, we have developed online remote radiation monitoring system to minimize number of entries in amber areas and to reduce the exposure to the surveyor and operator. This also helped in volume reduction of secondary waste. The reliability and accuracy of the online monitoring system is confirmed with actual measurements and by theoretical shielding calculations. The multiplication factors were estimated for remote online monitoring of Melter off Gas (MOG) filter, Hood filter, three exhaust filter banks, and overpack monitoring. This paper summarizes how the online remote monitoring system had helped in saving of 128.52 Person-mSv collective dose (14.28% of budgeted dose) and also there was 2.6 m 3 reduction in generation of Cat-I waste. (author)

  8. Conceptual design of a joule-heated ceramic melter for the DOE Fernald silos 1, 2, and 3 wastes

    International Nuclear Information System (INIS)

    Robinson, R.A.; Janke, D.S.; Peters, R.; Fekete, L.

    1992-06-01

    Vitrification of nuclear wastes has been under investigation since the mid-1950s. Most of the international communities experience has been with vitrification of high level nuclear wastes. In the US, this technology was developed by Battelle scientists at the DOEs Pacific Northwest Laboratories located at their Hanford site. Based on Laboratory and pilot-scale testing conducted at Hanford in the early 1970s, the DOE has constructed high level nuclear waste vitrification facilities at both Savannah River, South Carolina, and West Valley, New York, and is finalizing the design of a similar treatment facility at Hanford. Although these systems were designed to be fully remote due to the extreme radioactive hazards associated with this type of nuclear waste, technology transfer was successfully applied to the design of a vitrification process for the K-65 and uranium metal oxide wastes in a semi-remote operation at Fernald. This paper describes a conceptual design of a joule-heated, slurry-fed ceramic melter that was developed for vitrification of the DOE K-65 and metal oxide low level wastes at Fernald, Ohio

  9. Hanford Site Waste Management Plan

    International Nuclear Information System (INIS)

    1988-12-01

    The Hanford Site Waste Management Plan (HWMP) was prepared in accordance with the outline and format described in the US Department of Energy Orders. The HWMP presents the actions, schedules, and projected costs associated with the management and disposal of Hanford defense wastes, both radioactive and hazardous. The HWMP addresses the Waste Management Program. It does not include the Environmental Restoration Program, itself divided into the Environmental Restoration Remedial Action Program and the Decontamination and Decommissioning Program. The executive summary provides the basis for the plans, schedules, and costs within the scope of the Waste Management Program at Hanford. It summarizes fiscal year (FY) 1988 including the principal issues and the degree to which planned activities were accomplished. It further provides a forecast of FY 1989 including significant milestones. Section 1 provides general information for the Hanford Site including the organization and administration associated with the Waste Management Program and a description of the Site focusing on waste management operations. Section 2 and Section 3 describe radioactive and mixed waste management operations and hazardous waste management, respectively. Each section includes descriptions of the waste management systems and facilities, the characteristics of the wastes managed, and a discussion of the future direction of operations

  10. Safeguardability of the vitrification option for disposal of plutonium

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-05-01

    Safeguardability of the vitrification option for plutonium disposition is rather complex and there is no experience base in either domestic or international safeguards for this approach. In the present treaty regime between the US and the states of the former Soviet Union, bilaterial verifications are considered more likely with potential for a third-party verification of safeguards. There are serious technological limitations to applying conventional bulk handling facility safeguards techniques to achieve independent verification of plutonium in borosilicate glass. If vitrification is the final disposition option chosen, maintaining continuity of knowledge of plutonium in glass matrices, especially those containing boron and those spike with high-level wastes or {sup 137}Cs, is beyond the capability of present-day safeguards technologies and nondestructive assay techniques. The alternative to quantitative measurement of fissile content is to maintain continuity of knowledge through a combination of containment and surveillance, which is not the international norm for bulk handling facilities.

  11. 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

  12. Waste Encapsulation and Storage Facility (WESF) Dangerous Waste Training Plan (DWTP)

    International Nuclear Information System (INIS)

    SIMMONS, F.M.

    2000-01-01

    This Waste Encapsulation Storage Facility (WESF) Dangerous Waste Training Plan (DWTP) applies to personnel who perform work at, or in support of WESF. The plan, along with the names of personnel, may be given to a regulatory agency inspector upon request. General workers, subcontractors, or visiting personnel who have not been trained in the management of dangerous wastes must be accompanied by an individual who meets the requirements of this training plan. Dangerous waste management includes handling, treatment, storage, and/or disposal of dangerous and/or mixed waste. Dangerous waste management units covered by this plan include: less-than-90-day accumulation area(s); pool cells 1-8 and 12 storage units; and process cells A-G storage units. This training plan describes general requirements, worker categories, and provides course descriptions for operation of the WESF permitted miscellaneous storage units and the Less-than-90-Day Accumulation Areas

  13. Final Report: Vitrification of Inorganic Ion-Exchange Media, VSL-16R3710-1

    Energy Technology Data Exchange (ETDEWEB)

    Kot, Wing K. [The Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Pegg, Ian L. [The Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Brandys, Marek [The Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Penafiel, Miguel [The Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.

    2018-02-21

    One of the primary roles of waste pretreatment at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is to separate the majority of the radioactive components from the majority of the nonradioactive components in retrieved tank wastes, producing a high level waste (HLW) stream and a low activity waste (LAW) stream. This separation process is a key element in the overall strategy to reduce the volume of HLW that requires vitrification and subsequent disposal in a national deep geological repository for high level nuclear waste. After removal of the radioactive constituents, the LAW stream, which has a much larger volume but smaller fraction of radioactivity than the HLW stream, will be immobilized and disposed of in near surface facilities at the Hanford site.

  14. FY 2001 Hanford Waste Management Strategic Plan

    International Nuclear Information System (INIS)

    COLLINS, M.S.

    2001-01-01

    We are pleased to present the 2001 Hanford Waste Management Program Strategic Plan. This plan supports the newly developed U. S. Department of Energy Site outcomes strategy. The 2001 Plan reflects current and projected needs for Waste Management Program services in support of Hanford Site cleanup, and updates the objectives and actions using new waste stream oriented logic for the strategic goals: (1) waste treatment/processing, storage, and disposal; (2) interfaces; and (3) program excellence. Overall direction for the Program is provided by the Waste Management Division, Office of the Assistant Manager for Environmental Restoration and Waste Management, U. S. Department of Energy, Richland Operations Office. Fluor Hanford, Inc. is the operating contractor for the program. This Plan documents proactive strategies for planning and budgeting, with a major focus on helping meet regulatory commitments in a timely and efficient manner and concurrently assisting us in completing programs cheaper, better and quicker. Newly developed waste stream oriented logic was incorporated to clarify Site outcomes. External drivers, technology inputs, treatment/processing, storage and disposal strategies, and stream specific strategies are included for the six major waste types addressed in this Plan (low-level waste, mixed low-level waste, contact-handled transuranic waste, remote-handled transuranic waste, liquid waste, and cesium/strontium capsules). The key elements of the strategy are identification and quantification of the needs for waste management services, assessment of capabilities, and development of cost-effective actions to meet the needs and to continuously improve performance. Accomplishment of specific actions as set forth in the Plan depends on continued availability of the required resources and funding. The primary objectives of Plan are: (1) enhance the Waste Management Program to improve flexibility, become more holistic especially by implementing new

  15. 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

  16. La Hague Continuous Improvement Program: Enhancement of the Vitrification Throughput

    International Nuclear Information System (INIS)

    Petitjean, V.; De Vera, R.; Hollebecque, J.F.; Tronche, E.; Flament, T.; Pereira Mendes, F.; Prod'homme, A.

    2006-01-01

    The vitrification of high-level liquid waste produced from nuclear fuel reprocessing has been carried out industrially for over 25 years by AREVA/COGEMA, with two main objectives: containment of the long lived fission products and reduction of the final volume of waste. At the 'La Hague' plant, in the 'R7' and 'T7' facilities, vitrified waste is obtained by first evaporating and calcining the nitric acid feed solution-containing fission products in calciners. The product-named calcinate- is then fed together with glass frit into induction-heated metallic melters to produce the so-called R7/T7 glass, well known for its excellent containment properties. Both facilities are equipped with three processing lines. In the near future the increase of the fuel burn-up will influence the amount of fission product solutions to be processed at R7/T7. As a consequence, in order to prepare these changes, it is necessary to feed the calciner at higher flow-rates. Consistent and medium-term R and D programs led by CEA (French Atomic Energy Commission, the AREVA/COGEMA's R and D and R and T provider), AREVA/COGEMA (Industrial Operator) and AREVA/SGN (AREVA/COGEMA's Engineering), and associated to the industrial feed back of AREVA/COGEMA operations, have allowed continuous improvement of the process since 1998: - The efficiency and limitation of the equipment have been studied and solutions for technological improvements have been proposed whenever necessary, - The increase of the feeding flow-rate has been implemented on the improved CEA test rig (so called PEV, Evolutional Prototype of Vitrification) and adapted by AREVA/SGN for the La Hague plant using their modeling studies; the results obtained during this test confirmed the technological and industrial feasibility of the improvements achieved, - After all necessary improved equipments have been implemented in R7/T7 facilities, and a specific campaign has been performed on the R7 facility by AREVA/COGEMA. The flow-rate to the

  17. Closed system for bovine oocyte vitrification

    Directory of Open Access Journals (Sweden)

    Helena Ševelová

    2012-01-01

    Full Text Available The aim of our study was to develop a vitrification carrier for bovine oocyte cryopreservation. The carrier was to be cheap enough, elementary in its construction and meet contemporary requirements for a safe closed system. In a closed system, a cell is prevented from direct exposure to liquid nitrogen, thus minimizing the risk of cross-contamination. Furthermore, two questions regarding the proper vitrification technique were resolved: if it is necessary to partially denude the oocytes before the vitrification process or whether intact cumulus oocyte complexes should be frozen; and if it is more advantageous to preheat the vitrification solutions to female body temperature (39 °C or to keep them at room temperature. Our results show that it is better to partially denude the oocytes prior to vitrification because cryopreserved intact cumulus oocyte complexes often proved dark, non-homogeneous or fragmented cytoplasm after warming, with many of them having visibly widened perivitelline spaces or fractured zonae pellucidae as a result of extensive damage during vitrification. Consequently, intact cumulus oocyte complexes showed significantly lower numbers of cleavage stage embryos on Day 3 compared to partially denuded oocytes (7.4% and 26%, respectively. On the other hand, the survival rate and following development of fertilized oocytes in preheated vitrification solution were equal to results reached at room temperature conditions. In conclusion, results achieved with the newly developed carrier were comparable to previously published studies and therefore they could be recommended for common use.

  18. Constitutional studies in the palladium-rhodium-tellurium (-oxygen) system. A contribution to elucidate the behaviour of Pd, Rh and Te in the vitrification process of high-level waste concentrates (HLWC)

    International Nuclear Information System (INIS)

    Hartmann, T.

    1996-01-01

    In the vitrification process of high-level waste concentrates (HLWC) from the reprocessing of nuclear spent fuel elements, about 30 different elements have to be immobilized in a solid matrix consisting of an alkali borosilicate glass. Most of the waste oxides are dissolved in the alkali borosilicate melt and become structural elements of the glasses when cooled. This, however, applies only partly to the platinum metals Ru, which forms RuO 2 , and palladium and rhodium, which deposit as sparingly soluble and electrically conducting tellurides. This might considerably impair the technical process of HLWC vitrification. Therefore, constitutional studies on the Pd-Rh-Te system became necessary. The phase diagram of the Pd-Rh-Te ternary system at temperatures of 1150, 1100, 1050, 1000, 950, 900 and 750 C was determined under inertial conditions. Oxygen exerts a major influence on the system. Already under limited availability of oxygen, the rhodium contents of the solid solution phases α 1 and α 2 are clearly diminished. Rhodium of the phases becomes oxidized selectively. The three-phase field α 1 +α 2 +L is shifted to higher palladium and tellurium contents, even oxygen is available to a limited extend only. With the oxygen in the air, the extension of the three-phase space is reduced markedly. The complex process chemistry of Pf, Rh and Te during the vitrification can be described by the state of the Pd-Rh-Te ternary system after annealing in (air) oxygen for limited periods of time. (orig./MM) [de

  19. Standard guide for characterization of radioactive and/or hazardous wastes for thermal treatment

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide identifies methods to determine the physical and chemical characteristics of radioactive and/or hazardous wastes before a waste is processed at high temperatures, for example, vitrification into a homogeneous glass ,glass-ceramic, or ceramic waste form. This includes waste forms produced by ex-situ vitrification (ESV), in-situ vitrification (ISV), slagging, plasma-arc, hot-isostatic pressing (HIP) and/or cold-pressing and sintering technologies. Note that this guide does not specifically address high temperature waste treatment by incineration but several of the analyses described in this guide may be useful diagnostic methods to determine incinerator off-gas composition and concentrations. The characterization of the waste(s) recommended in this guide can be used to (1) choose and develop the appropriate thermal treatment methodology, (2) determine if waste pretreatment is needed prior to thermal treatment, (3) aid in development of thermal treatment process control, (4) develop surrogate wa...

  20. Waste minimization and pollution prevention awareness plan

    International Nuclear Information System (INIS)

    1991-01-01

    The purpose of this plan is to document the Lawrence Livermore National Laboratory (LLNL) Waste Minimization and Pollution Prevention Awareness Program. The plan specifies those activities and methods that are or will be employed to reduce the quantity and toxicity of wastes generated at the site. The intent of this plan is to respond to and comply with (DOE's) policy and guidelines concerning the need for pollution prevention. The Plan is composed of a LLNL Waste Minimization and Pollution Prevention Awareness Program Plan and, as attachments, Program- and Department-specific waste minimization plans. This format reflects the fact that waste minimization is considered a line management responsibility and is to be addressed by each of the Programs and Departments. 14 refs